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6. Provide sufficient information to
show that a genuine dispute exists with
the applicant regarding a material issue
of law or fact. This information must
include references to specific portions
of the application (including the
applicant’s environmental report and
safety report) that the requester/
petitioner disputes and the supporting
reasons for each dispute, or, if the
requester/petitioner believes the
application fails to contain information
on a relevant matter as required by law,
the identification of each failure and the
supporting reasons for the requester’s/
petitioner’s belief.
In addition, in accordance with 10
CFR 2.309(f)(2), contentions must be
based on documents or other
information available at the time the
petition is to be filed, such as the
application, supporting safety analysis
report, environmental report or other
supporting document filed by an
applicant or licensee, or otherwise
available to the petitioner. On issues
arising under the National
Environmental Policy Act, the
requester/petitioner shall file
contentions based on the applicant’s
environmental report. The requester/
petitioner may amend those contentions
or file new contentions if there are data
or conclusions in the NRC draft, or final
environmental impact statement,
environmental assessment, or any
supplements relating thereto, that differ
significantly from the data or
conclusions in the applicant’s
documents. Otherwise, contentions may
be amended or new contentions filed
after the initial filing only with leave of
the presiding officer.
Each contention shall be given a
separate numeric or alpha designation
within one of the following groups:
1. Technical—primarily concerns
issues relating to matters discussed or
referenced in the Safety Evaluation
Report for the proposed action.
2. Environmental—primarily concerns
issues relating to matters discussed or
referenced in the Environmental Report
for the proposed action.
3. Emergency Planning—primarily
concerns issues relating to matters
discussed or referenced in the
Emergency Plan as it relates to the
proposed action.
4. Physical Security—primarily
concerns issues relating to matters
discussed or referenced in the Physical
Security Plan as it relates to the
proposed action.
5. Miscellaneous—does not fall into
one of the categories outlined above.
If the requester/petitioner believes a
contention raises issues that cannot be
classified as primarily falling into one of
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these categories, the requester/petitioner
must set forth the contention and
supporting bases, in full, separately for
each category into which the requester/
petitioner asserts the contention
belongs, with a separate designation for
that category.
Requesters/petitioners should, when
possible, consult with each other in
preparing contentions and combine
similar subject matter concerns into a
joint contention, for which one of the
co-sponsoring requesters/petitioners is
designated the lead representative.
Further, in accordance with 10 CFR
2.309(f)(3), any requester/petitioner that
wishes to adopt a contention proposed
by another requester/petitioner must do
so, in accordance with the E-Filing rule,
within 10 days of the date the
contention is filed, and designate a
representative who shall have the
authority to act for the requester/
petitioner.
In accordance with 10 CFR 2.309(g),
a request for hearing and/or petition for
leave to intervene may also address the
selection of the hearing procedures,
taking into account the provisions of 10
CFR 2.310.
III. Further Information
Documents related to this action,
including the application for
amendment and supporting
documentation, are available
electronically at the NRC’s Electronic
Reading Room at http://www.nrc.gov/
reading-rm/adams.html. From this site,
you can access the NRC’s Agencywide
Document Access and Management
System (ADAMS), which provides text
and image files of NRC’s public
documents. The ADAMS accession
number for the document related to this
Notice is ML073090651, Redacted
Version of Amendment Request for
Processing UF6 in the CD Line Facility
at the NFS Site. If you do not have
access to ADAMS or if there are
problems in accessing the documents
located in ADAMS, contact the NRC
Public Document Room (PDR) Reference
staff at 1–800–397–4209, 301–415–4737,
or by e-mail to pdr@nrc.gov.
These documents may also be viewed
electronically on the public computers
located at the NRC’s PDR, O 1 F21, One
White Flint North, 11555 Rockville
Pike, Rockville, MD 20852. The PDR
reproduction contractor will copy
documents for a fee.
Dated at Rockville, Maryland, this 12th day
of December 2007.
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For the Nuclear Regulatory Commission.
Peter J. Habighorst,
Chief, Fuel Manufacturing Branch, Fuel
Facility Licensing Directorate, Division of Fuel
Cycle Safety and Safeguards, Office of
Nuclear Material Safety and Safeguards.
[FR Doc. E7–25406 Filed 12–28–07; 8:45 am]
BILLING CODE 7590–01–P
NUCLEAR REGULATORY
COMMISSION
Biweekly Notice; Applications and
Amendments to Facility Operating
Licenses Involving No Significant
Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the
Atomic Energy Act of 1954, as amended
(the Act), the U.S. Nuclear Regulatory
Commission (the Commission or NRC
staff) is publishing this regular biweekly
notice. The Act requires the
Commission publish notice of any
amendments issued, or proposed to be
issued and grants the Commission the
authority to issue and make
immediately effective any amendment
to an operating license upon a
determination by the Commission that
such amendment involves no significant
hazards consideration, notwithstanding
the pendency before the Commission of
a request for a hearing from any person.
This biweekly notice includes all
notices of amendments issued, or
proposed to be issued from December 6,
2007 to December 19, 2007. The last
biweekly notice was published on
December 18, 2007 (72 FR 71703).
Notice of Consideration of Issuance of
Amendments to Facility Operating
Licenses, Proposed no Significant
Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a
proposed determination that the
following amendment requests involve
no significant hazards consideration.
Under the Commission’s regulations in
10 CFR 50.92, this means that operation
of the facility in accordance with the
proposed amendment would not (1)
involve a significant increase in the
probability or consequences of an
accident previously evaluated; or (2)
create the possibility of a new or
different kind of accident from any
accident previously evaluated; or (3)
involve a significant reduction in a
margin of safety. The basis for this
proposed determination for each
amendment request is shown below.
The Commission is seeking public
comments on this proposed
determination. Any comments received
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within 30 days after the date of
publication of this notice will be
considered in making any final
determination. Within 60 days after the
date of publication of this notice, the
licensee may file a request for a hearing
with respect to issuance of the
amendment to the subject facility
operating license and any person whose
interest may be affected by this
proceeding and who wishes to
participate as a party in the proceeding
must file a written request for a hearing
and a petition for leave to intervene.
Normally, the Commission will not
issue the amendment until the
expiration of 60 days after the date of
publication of this notice. The
Commission may issue the license
amendment before expiration of the 60day period provided that its final
determination is that the amendment
involves no significant hazards
consideration. In addition, the
Commission may issue the amendment
prior to the expiration of the 30-day
comment period should circumstances
change during the 30-day comment
period such that failure to act in a
timely way would result, for example in
derating or shutdown of the facility.
Should the Commission take action
prior to the expiration of either the
comment period or the notice period, it
will publish in the Federal Register a
notice of issuance. Should the
Commission make a final No Significant
Hazards Consideration Determination,
any hearing will take place after
issuance. The Commission expects that
the need to take this action will occur
very infrequently.
Written comments may be submitted
by mail to the Chief, Rulemaking,
Directives and Editing Branch, Division
of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory
Commission, Washington, DC 20555–
0001, and should cite the publication
date and page number of this Federal
Register notice. Written comments may
also be delivered to Room 6D22, Two
White Flint North, 11545 Rockville
Pike, Rockville, Maryland, from 7:30
a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received
may be examined at the Commission’s
Public Document Room (PDR), located
at One White Flint North, Public File
Area O1F21, 11555 Rockville Pike (first
floor), Rockville, Maryland. The filing of
requests for a hearing and petitions for
leave to intervene is discussed below.
Within 60 days after the date of
publication of this notice, person(s) may
file a request for a hearing with respect
to issuance of the amendment to the
subject facility operating license and
any person whose interest may be
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affected by this proceeding and who
wishes to participate as a party in the
proceeding must file a written request
via electronic submission through the
NRC E-Filing system for a hearing and
a petition for leave to intervene.
Requests for a hearing and a petition for
leave to intervene shall be filed in
accordance with the Commission’s
‘‘Rules of Practice for Domestic
Licensing Proceedings’’ in 10 CFR part
2. Interested person(s) should consult a
current copy of 10 CFR 2.309, which is
available at the Commission’s PDR,
located at One White Flint North, Public
File Area 01F21, 11555 Rockville Pike
(first floor), Rockville, Maryland.
Publicly available records will be
accessible from the Agencywide
Documents Access and Management
System’s (ADAMS) Public Electronic
Reading Room on the Internet at the
NRC Web site, http://www.nrc.gov/
reading-rm/doc-collections/cfr/. If a
request for a hearing or petition for
leave to intervene is filed within 60
days, the Commission or a presiding
officer designated by the Commission or
by the Chief Administrative Judge of the
Atomic Safety and Licensing Board
Panel, will rule on the request and/or
petition; and the Secretary or the Chief
Administrative Judge of the Atomic
Safety and Licensing Board will issue a
notice of a hearing or an appropriate
order.
As required by 10 CFR 2.309, a
petition for leave to intervene shall set
forth with particularity the interest of
the petitioner in the proceeding, and
how that interest may be affected by the
results of the proceeding. The petition
should specifically explain the reasons
why intervention should be permitted
with particular reference to the
following general requirements: (1) The
name, address, and telephone number of
the requestor or petitioner; (2) the
nature of the requestor’s/petitioner’s
right under the Act to be made a party
to the proceeding; (3) the nature and
extent of the requestor’s/petitioner’s
property, financial, or other interest in
the proceeding; and (4) the possible
effect of any decision or order which
may be entered in the proceeding on the
requestor’s/petitioner’s interest. The
petition must also set forth the specific
contentions which the petitioner/
requestor seeks to have litigated at the
proceeding.
Each contention must consist of a
specific statement of the issue of law or
fact to be raised or controverted. In
addition, the petitioner/requestor shall
provide a brief explanation of the bases
for the contention and a concise
statement of the alleged facts or expert
opinion which support the contention
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and on which the petitioner/requestor
intends to rely in proving the contention
at the hearing. The petitioner/requestor
must also provide references to those
specific sources and documents of
which the petitioner is aware and on
which the petitioner/requestor intends
to rely to establish those facts or expert
opinion. The petition must include
sufficient information to show that a
genuine dispute exists with the
applicant on a material issue of law or
fact. Contentions shall be limited to
matters within the scope of the
amendment under consideration. The
contention must be one which, if
proven, would entitle the petitioner/
requestor to relief. A petitioner/
requestor who fails to satisfy these
requirements with respect to at least one
contention will not be permitted to
participate as a party.
Those permitted to intervene become
parties to the proceeding, subject to any
limitations in the order granting leave to
intervene, and have the opportunity to
participate fully in the conduct of the
hearing.
If a hearing is requested, and the
Commission has not made a final
determination on the issue of no
significant hazards consideration, the
Commission will make a final
determination on the issue of no
significant hazards consideration. The
final determination will serve to decide
when the hearing is held. If the final
determination is that the amendment
request involves no significant hazards
consideration, the Commission may
issue the amendment and make it
immediately effective, notwithstanding
the request for a hearing. Any hearing
held would take place after issuance of
the amendment. If the final
determination is that the amendment
request involves a significant hazards
consideration, any hearing held would
take place before the issuance of any
amendment.
A request for hearing or a petition for
leave to intervene must be filed in
accordance with the NRC E-Filing rule,
which the NRC promulgated in August
28, 2007 (72 FR 49139). The E-Filing
process requires participants to submit
and serve documents over the Internet
or in some cases to mail copies on
electronic storage media. Participants
may not submit paper copies of their
filings unless they seek a waiver in
accordance with the procedures
described below.
To comply with the procedural
requirements of E-Filing, at least five (5)
days prior to the filing deadline, the
petitioner/requestor must contact the
Office of the Secretary by e-mail at
HEARINGDOCKET@NRC.GOV, or by
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calling (301) 415–1677, to request (1) a
digital ID certificate, which allows the
participant (or its counsel or
representative) to digitally sign
documents and access the E-Submittal
server for any proceeding in which it is
participating; and/or (2) creation of an
electronic docket for the proceeding
(even in instances in which the
petitioner/requestor (or its counsel or
representative) already holds an NRCissued digital ID certificate). Each
petitioner/requestor will need to
download the Workplace Forms
ViewerTM to access the Electronic
Information Exchange (EIE), a
component of the E-Filing system. The
Workplace Forms ViewerTM is free and
is available at http://www.nrc.gov/sitehelp/e-submittals/install-viewer.html.
Information about applying for a digital
ID certificate is available on NRC’s
public Web site at http://www.nrc.gov/
site-help/e-submittals/applycertificates.html.
Once a petitioner/requestor has
obtained a digital ID certificate, had a
docket created, and downloaded the EIE
viewer, it can then submit a request for
hearing or petition for leave to
intervene. Submissions should be in
Portable Document Format (PDF) in
accordance with NRC guidance
available on the NRC public Web site at
http://www.nrc.gov/site-help/esubmittals.html. A filing is considered
complete at the time the filer submits its
documents through EIE. To be timely,
an electronic filing must be submitted to
the EIE system no later than 11:59 p.m.
Eastern Time on the due date. Upon
receipt of a transmission, the E-Filing
system time-stamps the document and
sends the submitted an e-mail notice
confirming receipt of the document. The
EIE system also distributes an e-mail
notice that provides access to the
document to the NRC Office of the
General Counsel and any others who
have advised the Office of the Secretary
that they wish to participate in the
proceeding, so that the filer need not
serve the documents on those
participants separately. Therefore,
applicants and other participants (or
their counsel or representative) must
apply for and receive a digital ID
certificate before a hearing request/
petition to intervene is filed so that they
can obtain access to the document via
the E-Filing system.
A person filing electronically may
seek assistance through the ‘‘Contact
Us’’ link located on the NRC Web site
at http://www.nrc.gov/site-help/esubmittals.html or by calling the NRC
technical help line, which is available
between 8:30 a.m. and 4:15 p.m.,
Eastern Time, Monday through Friday.
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The help line number is (800) 397–4209
or locally, (301) 415–4737.
Participants who believe that they
have a good cause for not submitting
documents electronically must file a
motion, in accordance with 10 CFR
2.302(g), with their initial paper filing
requesting authorization to continue to
submit documents in paper format.
Such filings must be submitted by: (1)
First class mail addressed to the Office
of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission,
Washington, DC 20555–0001, Attention:
Rulemaking and Adjudications Staff; or
(2) courier, express mail, or expedited
delivery service to the Office of the
Secretary, Sixteenth Floor, One White
Flint North, 11555 Rockville Pike,
Rockville, Maryland 20852, Attention:
Rulemaking and Adjudications Staff.
Participants filing a document in this
manner are responsible for serving the
document on all other participants.
Filing is considered complete by firstclass mail as of the time of deposit in
the mail, or by courier, express mail, or
expedited delivery service upon
depositing the document with the
provider of the service.
Non-timely requests and/or petitions
and contentions will not be entertained
absent a determination by the
Commission, the presiding officer, or
the Atomic Safety and Licensing Board
that the petition and/or request should
be granted and/or the contentions
should be admitted, based on a
balancing of the factors specified in 10
CFR 2.309(c)(1)(i)–(viii). To be timely,
filings must be submitted no later than
11:59 p.m. Eastern Time on the due
date.
Documents submitted in adjudicatory
proceedings will appear in NRC’s
electronic hearing docket which is
available to the public at http://
ehd.nrc.gov/EHD_Proceeding/home.asp,
unless excluded pursuant to an order of
the Commission, an Atomic Safety and
Licensing Board, or a Presiding Officer.
Participants are requested not to include
personal privacy information, such as
social security numbers, home
addresses, or home phone numbers in
their filings. With respect to copyrighted
works, except for limited excerpts that
serve the purpose of the adjudicatory
filings and would constitute a Fair Use
application, participants are requested
not to include copyrighted materials in
their submission.
For further details with respect to this
amendment action, see the application
for amendment which is available for
public inspection at the Commission’s
PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville
Pike (first floor), Rockville, Maryland.
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Publicly available records will be
accessible from the ADAMS Public
Electronic Reading Room on the Internet
at the NRC Web site, http://
www.nrc.gov/reading-rm/adams.html. If
you do not have access to ADAMS or if
there are problems in accessing the
documents located in ADAMS, contact
the PDR Reference staff at 1 (800) 397–
4209, (301) 415–4737 or by e-mail to
pdr@nrc.gov.
Carolina Power & Light Company,
Docket No. 50–261, H. B. Robinson
Steam Electric Plant, Unit No. 2,
Darlington County, South Carolina.
Date of amendment request:
November 19, 2007.
Description of amendment request:
The proposed amendment would make
administrative revisions to delete
requirements that are obsolete or
redundant, or correct and clarify the
typing and formatting of other
requirements. The proposed changes
will not result in changes to the plant
design or the procedural controls for the
operation, surveillance, or maintenance
of the plant.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Do the Proposed Changes Involve a
Significant Increase in the Probability or
Consequences of an Accident Previously
Evaluated?
No. The proposed changes do not involve
a significant increase in the probability or
consequences of an accident previously
evaluated. The proposed changes are
administrative. The changes delete obsolete
or redundant requirements, clarify existing
requirements, and correct typing and
formatting errors. There will be no resulting
changes to the plant design or procedural
controls. Therefore, the proposed changes do
not involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Do the Proposed Changes Create the
Possibility of a New or Different Kind of
Accident From Any Previously Evaluated?
No. The proposed changes do not create
the possibility of a new or different kind of
accident from any previously evaluated.
There are no physical changes being made to
the plant or to the manner in which the plant
is operated. Therefore, the changes do not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. Do the Proposed Changes Involve a
Significant Reduction in the Margin of
Safety?
No. The proposed changes do not involve
a significant reduction in the margin of
safety. There are no physical changes being
made to the plant or to the manner in which
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the plant is operated. The proposed changes
are administrative. The changes delete
obsolete or redundant requirements, clarify
existing requirements, and correct typing and
formatting errors. Therefore, the changes do
not involve a significant reduction in any
margin of safety for HBRSEP [H.B. Robinson
Steam Electric Plant], Unit No. 2.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: David T.
Conley, Associate General Counsel II—
Legal. Department, Progress Energy
Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Branch Chief: Thomas H. Boyce.
sroberts on PROD1PC70 with NOTICES
Duke Power Company LLC, Docket No.
50–369, McGuire Nuclear Station, Unit
1, Mecklenburg County, North Carolina.
Date of amendment request: February
21, 2007, as supplemented August 9,
2007.
Description of amendment request:
The proposed amendment would allow,
on a one-time basis, an extension of the
interval governing the conduct of the
Integrated Leak Rate Test (ILRT) for
McGuire Nuclear Station, Unit 1. The
proposed amendment would revise
administrative Technical Specification
(TS) 5.5.2, ‘‘Containment Leak Rate
Testing Program,’’ from the currently
approved 15-year interval (since the last
McGuire Nuclear Station, Unit 1, Type
A test) to a frequency encompassing the
end of the McGuire Nuclear Station,
Unit 1, End-of-Cycle 19 refueling outage
(approximately 6 months beyond the
present TS frequency).
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Involve a significant increase in the
probability or consequences of an accident
previously evaluated, or
2. Create the possibility of a new or
different kind of accident from any accident
previously evaluated, or
3. Involve a significant reduction in a
margin of safety.
First Standard
The proposed amendment will not involve
a significant increase in the probability or
consequences of an accident previously
evaluated. The proposed extension to the
Type A testing intervals cannot increase the
probability of an accident previously
evaluated since extension of the intervals is
not a physical plant modification that could
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alter the probability of accident occurrence,
nor is it an activity or modification by itself
that could lead to equipment failure or
accident initiation. The proposed extension
to the Type A testing intervals does not result
in a significant increase in the consequences
of an accident as documented in NUREG–
1493 [‘‘Performance-Based Containment
Leak-Test Program’’, NUREG–1493,
September 1995]. The NUREG notes that very
few potential containment leakage paths are
not identified by Type B and Type C tests.
It concludes that reducing the Type A testing
frequency to once per twenty years leads to
an imperceptible increase in risk. McGuire
[Nuclear Station, Unit 1 (McGuire Unit 1)]
provides a high degree of assurance through
testing and inspection that the containment
will not degrade in a manner detectable only
by Type A testing. Prior Type A tests for
McGuire Unit 1 identified containment
leakage within acceptance criteria, indicating
a very leak tight containment. Inspections
required by the ASME Code [American
Society of Mechanical Engineers (ASME),
Boiler and Pressure Vessel Code (Code)] are
also performed in order to identify
indications of containment degradation that
could affect leak tightness. Separately, Type
B and Type C testing, required by TS
[Technical Specification] identify any
containment opening from design
penetrations, such as valves, that would
otherwise be detected by a Type A test. These
factors establish that an extension to the
Type A test intervals will not represent a
significant increase in the consequences of an
accident.
Second Standard
The proposed amendments will not create
the possibility of a new or different kind of
accident from any accident previously
evaluated. The proposed revisions to the
McGuire TS add a one-time extension to the
current interval for Type A testing. The
current test interval of fifteen years, based on
past performance, would be extended on a
one-time basis to approximately fifteen and
a half years from the last Type A test. The
proposed extension to the Type A test
interval does not create the possibility of a
new or different type of accident since there
are no physical changes being made to the
plants and there are no changes to the
operation of the plants that could introduce
a new failure mode.
Third Standard
The proposed amendment will not involve
a significant reduction in a margin of safety.
The proposed revisions to the McGuire TS
add a one-time extension to the current
interval for Type A testing. The current test
interval of fifteen years, based on past
performance, would be extended on a onetime basis to approximately fifteen and a half
years from the last Type A test. The proposed
extension to Type A test intervals will not
significantly reduce the margin of safety. The
NUREG–1493 generic study of the effects of
extending containment leakage testing
intervals found that a twenty-year interval
resulted in an imperceptible increase in risk
to the public. NUREG–1493 found that,
generically, the design containment leakage
rate contributes about 0.1 percent of the
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overall risk and that decreasing the Type A
testing frequency would have a minimal
effect on this risk, since 95 percent of the
Type A detectable leakage paths would
already be detected by Type B and Type C
testing. Similar proposed changes have been
previously reviewed and approved by the
NRC, and they are applicable to McGuire.
Based upon the preceding discussion, Duke
Energy Corporation [Duke Power Company,
LLC] has concluded that the proposed
amendments do not involve a significant
hazards consideration.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Ms. Lisa F.
Vaughn, Associate General Counsel and
Managing Attorney, Duke Energy
Carolinas, LLC, 526 South Church
Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Evangelos C.
Marinos.
Entergy Gulf States, Inc., and Entergy
Operations, Inc., Docket No. 50–458,
River Bend Station, Unit 1, West
Feliciana Parish, Louisiana.
Date of amendment request:
November 15, 2007.
Description of amendment request:
The proposed change would relocate
Surveillance Requirement (SR) 3.8.3.6
from the Technical Specifications (TS)
to a licensee-controlled document. SR
3.8.3.6 requires the Emergency Diesel
Generator (EDG) Fuel Oil Storage Tanks
(FOSTs) to be drained, sediment
removed, and cleaned on a 10-year
interval.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the proposed change involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The FOSTs provide the storage for the EDG
fuel oil, assuring an adequate volume is
available for each EDG to operate for seven
days in the event of a loss of offsite power
concurrent with a loss of coolant accident.
The relocation of the SR to drain and clean
the FOSTs will not impact any of the
previously analyzed accidents. Sediment in
the tank, or failure to perform this SR, does
not necessarily result in an inoperable
storage tank. Fuel oil quantity and quality are
assured by other TS SRs which remain
unchanged. These SRs help ensure tank
sediment is minimized and ensure that any
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degradation of the tank wall surface that
results in a fuel oil volume reduction is
detected and corrected in a timely manner.
As a result, adequate controls exist to allow
relocation of this preventative maintenance
cleaning requirement to licensee controlled
documents.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed TS changes do not involve
the addition or modification of any plant
equipment. Also, the proposed change will
not alter the design configuration, or method
of operation of plant equipment beyond its
normal functional capabilities. The proposed
TS change does not create any new credible
failure mechanisms, malfunctions or accident
initiators.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any previously
evaluated.
3. Does the proposed change involve a
significant reduction in a margin of safety?
Response: No.
The proposed change does not alter or
exceed a design basis or safety limit. Diesel
generator fuel oil quantity and quality will
continue to be maintained within acceptable
limits of the TS to assure the ability of the
EDG to perform its intended function.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Terence A.
Burke, Associate General Council—
Nuclear Entergy Services, Inc., 1340
Echelon Parkway, Jackson, Mississippi
39213.
NRC Branch Chief: Thomas G. Hilt.
sroberts on PROD1PC70 with NOTICES
Entergy Operations, Inc., System Energy
Resources, Inc., South Mississippi
Electric Power Association, and Entergy
Mississippi, Inc., Docket No. 50–416,
Grand Gulf Nuclear Station, Unit 1,
Claiborne County, Mississippi.
Date of amendment request:
December 5, 2007.
Description of amendment request:
The proposed amendment would
change the Grand Gulf Nuclear Station,
Unit 1 (GGNS), Technical Specification
(TS) 5.6.5, ‘‘Core Operating Limits
Report (COLR),’’ to add a reference to an
analytical method that will be used to
determine core operating limits. The
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new reference, NEDC–33383P, ‘‘GEXL97
Correlation Applicable to ATRIUM–10
Fuel,’’ will allow Entergy Operations,
Inc. (Entergy) to use a Global Nuclear
Fuel (GNF) method to determine fuel
assembly critical power of AREVA
ATRIUM–10 fuel. GGNS currently
operates with a full core of ATRIUM–10
fuel. Entergy plans to use the GEXL97
correlation for GGNS operating Cycle 17
currently scheduled to begin in the fall
2008. Additionally, an administrative
change is proposed to an existing
reference in TS 5.6.5.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the proposed change involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
Core operating limits are established each
operating cycle in accordance with TS 3.2,
‘‘Power Distribution’’ and TS 5.6.5, ‘‘Core
Operating Limits Report (COLR)’’. These core
operating limits ensure that the fuel design
limits are not exceeded during any
conditions of normal operation or in the
event of any Anticipated Operational
Occurrence (AOO). The methods used to
determine the operating limits are those
previously found acceptable by the NRC and
listed in TS Section 5.6.5.b.
A change to TS 5.6.5.b is requested to
include an additional reference to the list of
analytical methods. GGNS currently operates
with a full core of AREVA ATRIUM–10 fuel
but is scheduled to load GE14 fuel during the
next refueling outage. GGNS plans to use the
analysis methods of the new fuel vendor,
GNF for the analysis of the mixed core. The
GEXL97 correlation accurately models
predicted core behavior and appropriately
determines the overall critical power
uncertainty of the method. In addition, the
GEXL97 application range covers the range of
expected operation of the ATRIUM–10 fuel
during normal steady state and transient
conditions in the GGNS reload cores.
Although a depressurization transient could
result in vessel pressures below the range of
GEXL97, the transient would not threaten
fuel cladding integrity, since the margin to
the MCPR [minimum critical power ratio]
safety limit increases with decreasing reactor
pressure.
Additionally, Entergy proposes an
administrative change to the GESTAR-Il
reference in TS 5.6.5.b. The administrative
change does not alter any method of analysis
as described in the NRC approved versions
of GESTAR-II. The requested TS changes
concern the use of analytical methods and do
not involve any plant modifications or
operational changes that could affect any
postulated accident precursors or accident
mitigation systems and do not introduce any
new accident initiation mechanisms. The
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proposed changes have no effect on the type
or amount of radiation released, and have no
effect on predicted offsite doses in the event
of an accident. Thus, the proposed change
does not affect the probability of an accident
previously evaluated nor does it increase the
radiological consequences of any accident
previously evaluated.
Therefore, the proposed change does not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed TS changes will not change
the design function, reliability, performance,
or operation of any plant systems,
components, or structures. It does not create
the possibility of a new failure mechanism,
malfunction, or accident initiators not
considered in the design and licensing bases.
Plant operation will continue to be within
the core operating limits that are established
using NRC approved methods that are
applicable to the GGNS design and the GGNS
fuel.
Therefore, the proposed change does not
create the possibility of a new or different
kind of accident from any previously
evaluated.
3. Does the proposed change involve a
significant reduction in a margin of safety?
Response: No.
The proposed change adds GEXL97 to the
list of analytical methods in TS 5.6.5.b that
can be used to determine core operating
limits. Use of the GEXL97 correlation
analytical method provides an equivalent
level of protection as that currently provided.
The administrative change does not alter any
method of analysis as described in the NRC
approved versions of GESTAR-II. The
proposed change does not modify the safety
limits or set points at which protective
actions are initiated, and does not change the
requirements governing operation or
availability of safety equipment assumed to
operate to preserve the margin of safety.
Therefore, the proposed change does not
involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Terence A.
Burke, Associate General Council—
Nuclear Entergy Services, Inc., 1340
Echelon Parkway, Jackson, Mississippi
39213.
NRC Branch Chief: Thomas G. Hiltz.
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Exelon Generation Company, LLC,
Docket No. 50–353, Limerick Generating
Station, Unit 2, Montgomery County,
Pennsylvania.
sroberts on PROD1PC70 with NOTICES
Date of amendment request:
November 16, 2007.
Description of amendment request:
The proposed changes revise technical
specification (TS) action requirements
associated with inoperable reactor
coolant system (RCS) leakage detection
systems. A new TS action requirement
is proposed that will address the
inoperability of the drywell unit cooler
condensate flow rate monitoring system
concurrent with one other RCS leakage
detection system, other than the primary
containment atmosphere gaseous
radioactivity monitoring system. This
would relax the allowed out-of-service
time for the specified combination of
systems and is related to the current
inoperability of the drywell unit cooler
condensate flow rate monitoring system.
The proposed changes would be
effective for the remainder of the current
operating cycle (Cycle 10), which is
currently scheduled to end in the spring
of 2009, or until the next shutdown of
sufficient duration to allow for drywell
unit cooler condensate flow rate
monitoring system repairs, whichever
comes first.
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Do the proposed changes involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed changes continue to
maintain an acceptable level of reactor
coolant system (RCS) leakage detection
instrumentation required to support plant
operations. The level of RCS leakage
detection capability inherent with the
proposed changes will continue to provide
acceptable early warning detection of
potential RCS pressure boundary
degradation. The proposed changes do not
impact the physical configuration or design
function of plant structures, systems, or
components (SSCs) or the manner in which
SSCs are operated, modified, tested, or
inspected [with the exception of an increase
in allowed out-of-service time for a
concurrent inoperability of the drywell unit
cooler condensate flow rate monitoring
system and another specified RCS leakage
detection system]. The proposed changes do
not impact the initiators or assumptions of
analyzed events, nor do they impact
mitigation of accidents or transient events.
Therefore, the proposed changes do not
involve a significant increase in the
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20:08 Dec 28, 2007
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probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed changes only affect systems
associated with the detection of leakage
resulting from the degradation of the RCS
pressure boundary. The proposed changes do
not alter plant configuration or require that
new plant equipment be installed. The RCS
leakage detection systems will continue to
function as designed in all modes of
operation. No new accident type is created as
a result of the proposed changes. No new
failure mode for any equipment is created.
The proposed changes do not alter
assumptions made about accidents
previously evaluated. Therefore, the
proposed changes do not create the
possibility of a new or different kind of
accident from any accident previously
evaluated.
3. Do the proposed changes involve a
significant reduction in a margin of safety?
Response: No.
The proposed changes do not involve any
physical changes to plant SSCs or the manner
in which SSCs are operated, modified, tested,
or inspected. The proposed changes do not
involve a change to any safety limits, limiting
safety system settings, limiting conditions of
operation, or design parameters for any SSC.
The proposed changes do not impact any
safety analysis assumptions and do not
involve a change in initial conditions, system
response times, or other parameters affecting
an accident analysis. Therefore, the proposed
changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, with changes as noted above, it
appears that the three standards of 10
CFR 50.92(c) are satisfied. Therefore, the
NRC staff proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: J. Bradley
Fewell, Esquire, Associate General
Counsel, Exelon Generation Company,
LLC, 4300 Winfield Road, Warrenville,
IL 60555.
NRC Branch Chief: Harold K.
Chernoff.
Exelon Generation Company, LLC,
Docket Nos. 50–254 and 50–265, Quad
Cities Nuclear Power Station, Units 1
and 2, Rock Island County, Illinois.
Date of amendment request: October
9, 2007.
Description of amendment request: A
change is proposed to the technical
specifications (TS) of Quad Cities
Nuclear Power Station (QCNPS), Units 1
and 2, consistent with Technical
Specifications Task Force (TSTF)
Change Traveler TSTF–423 to the
standard TSs for boiling water reactor
plants, to allow, for some systems, entry
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74359
into hot shutdown rather than cold
shutdown to repair equipment, if risk is
assessed and managed consistent with
the program in place for complying with
the requirements of Title 10 of the Code
of Federal Regulations (10 CFR) Section
50.65(a)(4). Changes proposed herein
will be made to the QCNPS, Units 1 and
2, TSs for selected required action end
states providing this allowance.
The licensee reviewed the proposed
no significant hazards consideration
(NSHC) determination published in the
Federal Register on March 23, 2007 (71
FR 14726) and concluded that it is
applicable to QCNPS, Units 1 and 2.
The licensee incorporated the proposed
determination by reference to satisfy the
requirements of 10 CFR 50.91(a).
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), an
analysis of the issue of NSHC is
presented below:
1. The Proposed Change Does Not Involve
a Significant Increase in the Probability or
Consequences of an Accident Previously
Evaluated.
The proposed change allows a change to
certain required end states when the TS
Completion Times for remaining in power
operation will be exceeded. Most of the
requested technical specification (TS)
changes are to permit an end state of hot
shutdown (Mode 3) rather than an end state
of cold shutdown (Mode 4) contained in the
current TS. The request was limited to: (1)
Those end states where entry into the
shutdown mode is for a short interval, (2)
entry is initiated by inoperability of a single
train of equipment or a restriction on a plant
operational parameter, unless otherwise
stated in the applicable technical
specification, and (3) the primary purpose is
to correct the initiating condition and return
to power operation as soon as is practical.
Risk insights from both the qualitative and
quantitative risk assessments were used in
specific TS assessments. Such assessments
are documented in Section 6 of GE NEDC–
32988, Revision 2, ‘‘Technical Justification to
Support Risk Informed Modification to
Selected Required Action End States for BWR
Plants.’’ They provide an integrated
discussion of deterministic and probabilistic
issues, focusing on specific technical
specifications, which are used to support the
proposed TS end state and associated
restrictions. The staff finds that the risk
insights support the conclusions of the
specific TS assessments. Therefore, the
probability of an accident previously
evaluated is not significantly increased, if at
all. The consequences of an accident after
adopting proposed TSTF–423, are no
different than the consequences of an
accident prior to adopting TSTF–423.
Therefore, the consequences of an accident
previously evaluated are not significantly
affected by this change. The addition of a
requirement to assess and manage the risk
introduced by this change will further
minimize possible concerns. Therefore, this
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change does not involve a significant
increase in the probability or consequences
of an accident previously evaluated.
2. The Proposed Change Does Not Create
the Possibility of a New or Different Kind of
Accident From Any Previously Evaluated.
The proposed change does not involve a
physical alteration of the plant (no new or
different type of equipment will be installed).
If risk is assessed and managed, allowing a
change to certain required end states when
the TS Completion Times for remaining in
power operation are exceeded, i.e., entry into
hot shutdown rather than cold shutdown to
repair equipment, will not introduce new
failure modes or effects and will not, in the
absence of other unrelated failures, lead to an
accident whose consequences exceed the
consequences of accidents previously
evaluated. The addition of a requirement to
assess and manage the risk introduced by this
change and the commitment by the licensee
to adhere to the guidance in TSTF–IG–05–02,
Implementation Guidance for TSTF–423,
Revision 0, ‘‘Technical Specifications End
States, NEDC–32988–A,’’ will further
minimize possible concerns. Thus, this
change does not create the possibility of a
new or different kind of accident from an
accident previously evaluated.
3. The Proposed Change Does Not Involve
a Significant Reduction in the Margin of
Safety.
The proposed change allows, for some
systems, entry into hot shutdown rather than
cold shutdown to repair equipment, if risk is
assessed and managed. The BWROG’s
[Boiling Water Reactor Owners Group’s] risk
assessment approach is comprehensive and
follows staff guidance as documented in RGs
[Regulatory Guides] 1.174 and 1.177. In
addition, the analyses show that the criteria
of the three-tiered approach for allowing TS
changes are met. The risk impact of the
proposed TS changes was assessed following
the three-tiered approach recommended in
RG 1.177. A risk assessment was performed
to justify the proposed TS changes. The net
change to the margin of safety is
insignificant. Therefore, this change does not
involve a significant reduction in a margin of
safety.
sroberts on PROD1PC70 with NOTICES
Therefore, the NRC staff proposes to
determine that the requested
amendments involve no significant
hazards consideration.
Attorney for licensee: Mr. Bradley J.
Fewell, Associate General Counsel,
Exelon Generation Company, LLC, 4300
Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
Luminant Generation Company LLC,
Docket Nos. 50–445 and 50–446,
Comanche Peak Steam Electric Station,
Units 1 and 2, Somervell County, Texas
Date of amendment request:
November 29, 2007.
Brief description of amendments:
Revision to Technical Specification (TS)
3.6.7, (‘‘Spray Additive System,’’ to
allow modifications to the facility
potentially required to comply with U.S.
Nuclear Regulatory Commission (NRC)
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20:08 Dec 28, 2007
Jkt 214001
Generic Letter 2004–02, ‘‘Potential
Impact of Debris Blockage on
Emergency Recirculation during Design
Basis Accident at Pressurized Water
Reactors.’’
Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Do the proposed changes involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No
The proposed change[s] [do] not impact the
initiation or probability of occurrence of any
accident.
The accidents evaluated in the Final Safety
Analysis report (FSAR) that could be affected
by this proposed change are those involving
the pressurization of the containment and
those involving recirculation of fluid within
the Emergency Core Cooling System (ECCS)
or the Containment Spray System (e.g., loss
of coolant accidents (LOCAs)).
The change to a minimum pH [potential of
Hydrogen] of 7.1 will not result in a
significant increase in the radiological
consequences of a LOCA as-described below.
The equilibrium spray pH during the
recirculation phase resulting from this
change will be greater than or equal to 7.1.
The pH range for the spray will be bounded
by the water spray solution which is borated
water with a maximum of 2600 parts per
million (ppm) boron buffered to a final spray
solution pH much less than the 10.5 as
described in the current FSAR Section
3.11(B) for the postulated spray solution
environment. The maximum pH is the
limiting parameter for equipment
qualification. Since the resulting pH level
will be closer to neutral using the lower limit
of 7.1, post-LOCA corrosion of containment
components will not be increased. PostLOCA hydrogen generation will be reduced.
There will not be an adverse radiation dose
effect on any safety-related equipment. Thus,
the potential for failures of the ECCS or
safety-related equipment following a LOCA
will not be increased as a result of the
proposed change.
This modification affects the Containment
Spray System which is intended to respond
to and mitigate the effects of a LOCA. The
Containment Spray System will continue to
function in a manner consistent with the
plant design basis. There will be no
degradation in the performance of nor an
increase in the number of challenges to
equipment assumed to function during an
accident situation.
Therefore, these Technical Specification
(TS) revisions do not affect the probability of
any event initiators. There will be no adverse
changes to normal plant operating
parameters, Engineered Safety Features (ESF)
actuation setpoints, or accident mitigation
capabilities.
The proposed change allows the Spray
Additive System currently used to mitigate
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the consequences of an accident to maintain
the equilibrium sump pH at greater than or
equal to 7.1 to minimize chloride-induced
stress corrosion cracking in austenitic
stainless components important to safety
located inside containment. Therefore, the
proposed changes will not increase the
probability of an accident or malfunction of
equipment important to safety previously
evaluated in the FSAR.
The offsite and control room doses will
continue to meet the requirements of [Title
10 of the Code of Federal Regulations (10
CFR) part 100] 10 CFR 100, 10 CFR 50
Appendix A [General Design Criterion] GDC
19, [Standard Review Plan] SRP 15.6.5.11,
and SRP 6.4.11. The proposed new pH limit
will provide satisfactory retention of iodine
in the sump water, as well as provide
adequate pH control to minimize the
potential of chloride-induced stress corrosion
cracking of austenitic stainless steel
components.
Therefore, the proposed changes do not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The proposed change to the revised
Surveillance for the Containment Spray
Additive System provides for a required
minimum equilibrium pH in containment
post accident. There are no electrical or
mechanical components being added whose
failure could prevent the system from
functioning.
No new accident scenarios, transient
precursors, or limiting single failures are
introduced as a result of the proposed
changes. There will be no adverse effect or
challenges imposed on any safety-related
system as a result of this proposed change.
The amount of sodium hydroxide (NaOH)
will provide a minimum equilibrium sump
pH of 7.1 following mixing. Therefore, the
possibility of a new or different type of
accident is not created.
There are no changes which would cause
the malfunction of safety-related equipment,
assumed to be operable in the accident
analyses, as a result of the proposed
Technical Specification changes. The
possibility of a malfunction of safety-related
equipment with a different result is not
created.
Therefore, the proposed change[s] [do] not
create the possibility of a new or different
kind of accident from any previously
evaluated.
3. Do the proposed changes involve a
significant reduction in a margin of safety?
Response: No
The only function of the chemical additive
system is to provide pH control of the postaccident containment recirculation sump
water, since the borated water from the
Refueling Water Storage Tank (RWST) used
as the containment spray pump suction
source during injection is sufficient to
remove iodine from the containment
atmosphere following a LOCA. The net effect
on the pH control function of reducing the
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amount of buffer is that the equilibrium
sump pH will be lowered to a minimum of
7.1. There will be no change to the current
Technical Specification acceptance limits on
RWST volume and boron concentration. The
resulting equilibrium sump pH level from
this change will be closer to neutral;
therefore, the post-LOCA corrosion of
containment components will not be
increased (i.e., would be reduced).
Because the long term pH will be
maintained greater than or equal to 7.1,
margin to minimize the potential for stress
corrosion cracking is maintained.
The radiological analysis, as discussed in
the technical analysis above, is shown not to
be impacted. There will be no change to the
[departure from nucleate boiling ratio] DNBR
Correlation Limit, the design DNBR limits, or
the safety analysis DNBR limits discussed in
Bases Section 2.1.1. There will be no effect
on the manner in which Safety Limits or
Limiting Safety System Settings are
determined nor will there be any effect on
those plant systems necessary to assure the
accomplishment of protection functions.
There will be no adverse impact on
Departure of Nucleate Boiling Ratio limits,
[heat flux hot channel factor] FQ, [nuclear
enthalpy rise hot channel factor] F-delta-H,
LOCA peak cladding temperature, peak local
power density, or any other margin of safety.
Therefore the proposed change[s] [do] not
involve a reduction in a margin of safety.
sroberts on PROD1PC70 with NOTICES
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Timothy P.
Matthews, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW.,
Washington, DC 20036.
NRC Branch Chief: Thomas G. Hiltz.
Pacific Gas and Electric Company,
Docket Nos. 50–275 and 50–323, Diablo
Canyon Nuclear Power Plant, Unit Nos.
1 and 2, San Luis Obispo County,
California.
Date of amendment requests: October
2, 2007.
Description of amendment requests:
The proposed amendments would
revise Technical Specification (TS)
3.5.4, ‘‘Refueling Water Storage Tank
(RWST),’’ Surveillance Requirement
(SR) 3.5.4.2, to increase the minimum
required borated water volume from ‘‘≥
[greater than or equal to] 400,000
gallons (81.5% indicated level)’’ to ‘‘≥
455,300 gallons (93.6% level),’’ to
reflect the new sump design required to
comply with U.S. Nuclear Regulatory
Commission (NRC) Generic Letter 2004–
02, ‘‘Potential Impact of Debris Blockage
on Emergency Recirculation during
Design-Basis Accident at Pressurized
Water Reactors.’’
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Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. [Do] the proposed change[s] involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
The proposed change[s] [revise] the
minimum RWST borated water volume. The
RWST borated water volume is not an
initiator of any accident previously
evaluated. As a result, the probability of an
accident previously evaluated is not affected.
The proposed change[s] [do] not alter or
prevent the ability of structures, systems, and
components from performing their intended
function to mitigate the consequences of an
initiating event within the assumed
acceptance limits. The effect on containment
flood level, equipment qualification, and
containment sump pH remain within the
limits assumed in the design and accident
analyses. The proposed change[s] [do] not
affect the source term, containment isolation,
or radiological release assumptions used in
evaluating the radiological consequences of
an accident previously evaluated. Further,
the proposed change[s] [do] not increase the
types or amounts of radioactive effluent that
may be released offsite, nor significantly
increase individual or cumulative
occupational/public radiation exposures. The
proposed change[s] are consistent with the
safety analysis assumptions and resultant
consequences.
Therefore, the proposed change[s] [do] not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. [Do] the proposed change[s] create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
The change[s] [do] not involve a physical
alteration of the plant (i.e., no new or
different components or physical changes are
involved with this change) or a change in the
methods governing normal plant operation.
The change[s] [do] not alter any assumptions
made in the safety analysis.
Therefore, the proposed change[s] will not
create the possibility of a new or different
kind of accident from any accident
previously evaluated.
3. [Do] the proposed change[s] involve a
significant reduction in a margin of safety?
Response: No.
The proposed change[s] to revise the
required RWST minimum borated water
volume [do] not alter the manner in which
safety limits, limiting safety system settings
or limiting conditions for operation are
determined. The safety analysis acceptance
criteria are not affected by [these] change[s].
The proposed change[s] will not result in
plant operation in a configuration outside of
the design basis.
Therefore, the proposed change[s] [do] not
involve a significant reduction in a margin of
safety.
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The NRC staff has reviewed the licensee’s
analysis and, based on this review, it appears
that the three standards of 10 CFR 50.92(c)
are satisfied. Therefore, the NRC staff
proposes to determine that the amendment
requests involve no significant hazards
consideration.
Attorney for licensee: Jennifer Post, Esq.,
Pacific Gas and Electric Company, P.O. Box
7442, San Francisco, California 94120.
NRC Branch Chief: Thomas G. Hiltz.
Pacific Gas and Electric Co., Docket No. 50–
133, Humboldt Bay Power Plant (HBPP), Unit
3 Humboldt County, California.
Date of amendment request: November 5,
2007.
Description of amendment request: The
licensee has proposed amending the
technical specifications (TS) to delete many
operational and administrative requirements
upon transfer of spent nuclear fuel
assemblies and fuel fragment containers from
the Spent Fuel Pool (SFP) to the Humboldt
Bay Independent Spent Fuel Storage
Installation (ISFSI). Some TS requirements
will be relocated to the HBPP Quality
Assurance Plan.
Basis for proposed no significant hazards
consideration determination: As required by
10 CFR 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the change involve a significant
increase in the probability or consequences
of an accident previously evaluated?
Response: No.
The proposed changes reflect the transfer
of spent fuel from the Spent Fuel Pool to the
Humboldt Bay (HB) Independent Spent Fuel
Storage Installation. Design basis accidents
related to the SFP are discussed in the
Humboldt Bay Power Plant Unit 3 Defueled
Safety Analysis Report (DSAR). These
postulated accidents are predicated on spent
fuel being stored in the SFP. With the
removal of the spent fuel from the SFP, there
are no important-to-safety systems, structures
or components required to function or to be
monitored. In addition, there are no
remaining credible accidents involving spent
fuel or the SFP that require actions of a
Certified Fuel Handler or Noncertified Fuel
Handler to prevent occurrence or to mitigate
consequences. The proposed change to the
Design Features section of the Technical
Specifications (TS) clarifies that the spent
fuel is being stored in dry casks within an
ISFSI. The probability or consequences of
accidents at the ISFSI are evaluated in the HB
ISFSI Final Safety Analysis Report (FSAR)
and are independent of the accidents
evaluated in the HBPP Unit 3 DSAR.
Therefore, the proposed changes will not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
(2) Does the change create the possibility
of a new or different kind of accident from
any accident evaluated?
Response: No.
The proposed changes reflect the reduced
operational risks as a result of the spent fuel
being transferred to dry casks within an
ISFSI. The proposed changes do not modify
any systems, structures or components. The
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plant conditions for which the HBPP Unit 3
DSAR design basis accidents relating to spent
fuel and the SFP have been evaluated are no
longer applicable. The aforementioned
proposed changes do not affect any of the
parameters or conditions that could
contribute to the initiation of an accident.
Design basis accidents associated with the
dry cask storage of spent fuel are already
considered in the HB ISFSI FSAR. No new
accident scenarios are created as a result of
deleting nonapplicable operational and
administrative requirements. Therefore, the
proposed changes will not create the
possibility of a new or different kind of
accident from those previously evaluated.
(3) Does the change involve a significant
reduction in a margin of safety?
Response: No.
The proposed changes reflect the reduced
operational risks as a result of the spent fuel
being transferred to dry casks within an
ISFSI. The design basis and accident
assumptions within the HBPP Unit 3 DSAR
and the TS relating to spent fuel are no
longer applicable. The proposed changes do
not affect remaining plant operations, nor
structures, systems, or components
supporting decommissioning activities. In
addition, the proposed changes do not result
in a change in initial conditions, system
response time, or in any other parameter
affecting the course of a decommissioning
activity accident analysis. Therefore, the
proposed changes will not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: Ms. Jennifer K.
Post, Pacific Gas and Electric Company,
77 Beale Street, B30A, San Francisco,
CA.
NRC Branch Chief: Andrew Persinko.
sroberts on PROD1PC70 with NOTICES
Union Electric Company, Docket No.
50–483, Callaway Plant, Unit 1,
Callaway County, Missouri.
Date of amendment request: October
31, 2007.
Description of amendment request:
The amendment would revise Technical
Specification (TS) 3.8.1, ‘‘Essential
Service Water System (ESW),’’ and TS
3.8.1, ‘‘AC [Alternating Current]
Sources—Operating.’’ A note would be
added to Condition A, one ESW train
inoperable, of TS 3.8.1, and Condition
B, one diesel generator (DG) inoperable,
of TS 3.8.1 would be revised. The
revisions are to allow a one-time
completion time extension from 72
hours to 14 days to support a planned
replacement of ESW piping prior to
December 31, 2008, in the licensee’s fall
2008 refueling outage.
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Basis for proposed no significant
hazards consideration determination:
As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the
issue of no significant hazards
consideration, which is presented
below:
1. Does the proposed change involve a
significant increase in the probability or
consequences of an accident previously
evaluated?
Response: No.
[The only change to the plant is that
existing ESW piping will be replaced in the
fall 2008 refueling outage. There are no other
changes to the plant and no hardware or
equipment will be added to the plant. This
replacement is to address localized
degradation of the ESW piping due to
microbiologically induced corrosion.]
Overall protection system performance will
remain within the bounds of the previously
performed accident analyses since no
hardware changes are proposed to the
protection systems. The same reactor trip
system (RTS) and engineered safety feature
actuation system (ESFAS) instrumentation
will continue to be used. The protection
systems will continue to function in a
manner consistent with the plant design
basis. The use of polyethylene (PE) piping
[(i.e., replacing existing ESW piping by PE
piping)] in the ESW system in accordance
with ASME [American Society of Mechanical
Engineers Boiler and Pressure Vessel Code]
Code Case N–755, with justified materials
and design exceptions as noted in [the
licensee’s letter dated August 30, 2007
(ULNRC–05434), which requested relief from
the ASME Code to replace the ESW piping
by the PE piping], will [have the PE piping
that replaces the ESW piping] provide an
acceptable level of quality and safety. There
will be no changes to the essential service
water (ESW) system or [the] ultimate heat
sink (UHS) surveillance and operating limits.
[The licensee’s letter dated August 30, 2007,]
demonstrates the acceptability of using the
PE piping in this buried ASME Class 3
application [(i.e., replacing existing ESW
piping)].
The proposed changes will not adversely
affect accident initiators or precursors nor
alter the design assumptions, conditions, and
configurations of the facility or the manner
in which the plant is operated and
maintained. The proposed changes will not
alter or prevent the ability of structures,
systems, and components (SSCs) from
performing their intended [safety] functions
to mitigate the consequences of an initiating
event within the assumed acceptance limits.
The proposed changes do not affect the
way in which safety-related systems perform
their [safety] functions.
All accident analysis acceptance criteria
will continue to be met with the proposed
changes. The proposed changes will not
affect the source term, containment isolation,
or radiological release assumptions used in
evaluating the radiological consequences of
an accident previously evaluated. The
proposed changes will not alter any
assumptions or change any mitigation actions
in the radiological consequence evaluations
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Fmt 4703
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in the FSAR [Final Safety Analysis Report for
the Callaway Plant].
The applicable radiological dose
acceptance criteria [is unchanged] and will
continue to be met.
Therefore, the proposed changes do not
involve a significant increase in the
probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the
possibility of a new or different kind of
accident from any accident previously
evaluated?
Response: No.
There are no proposed changes in the
method by which any safety-related plant
SSC performs its safety function. [The
proposed changes will not affect the
performance of the ESW piping in terms of
providing mitigation of design basis
accidents per the FSAR accident analyses.]
The proposed changes will not affect the
normal method of plant operation or change
any operating parameters. No equipment
performance requirements will be affected.
The proposed changes will not alter any
assumptions made in the safety analyses.
No new accident scenarios, transient
precursors, failure mechanisms, or limiting
single failures will be introduced as a result
of this amendment. There will be no adverse
effect or challenges imposed on any safetyrelated system as a result of this amendment.
The proposed amendment will not alter the
design or performance of the 7300 Process
Protection System, Nuclear Instrumentation
System, or Solid State Protection System
used in the plant protection systems.
Therefore, the proposed changes do not
create the possibility of a new or different
[kind of] accident from any accident
previously evaluated.
3. Does the proposed change involve a
significant reduction in a margin of safety?
Response: No.
There will be no effect on those plant
systems necessary to assure the
accomplishment of protection functions.
There will be no impact on the overpower
limit, departure from nucleate boiling ratio
(DNBR) limits, heat flux hot channel factor
(FQ), nuclear enthalpy rise hot channel factor
(F [delta] H), loss of coolant accident peak
cladding temperature (LOCA PCT), peak
local power density, or any other margin of
safety. The applicable radiological dose
consequence acceptance criteria will
continue to be met. [The proposed changes
will not affect the performance of the ESW
piping in terms of providing mitigation of
design basis accidents per the FSAR accident
analyses.]
The proposed changes do not eliminate
any surveillances or alter the frequency of
[any] surveillances required by the Technical
Specifications. None of the acceptance
criteria for any accident analyses will be
changed.
The proposed changes will have no impact
on the radiological consequences of a design
basis accident.
Therefore, the proposed changes do not
involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the
licensee’s analysis and, based on this
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review, it appears that the three
standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff
proposes to determine that the
amendment request involves no
significant hazards consideration.
Attorney for licensee: John O’Neill,
Esq., Pillsbury Winthrop Shaw Pittman
LLP, 2300 N Street, NW., Washington,
DC 20037.
NRC Branch Chief: Thomas G. Hiltz.
Previously Published Notices of
Consideration of Issuance of
Amendments to Facility Operating
Licenses, Proposed no Significant
Hazards Consideration Determination,
and Opportunity for a Hearing
The following notices were previously
published as separate individual
notices. The notice content was the
same as above. They were published as
individual notices either because time
did not allow the Commission to wait
for this biweekly notice or because the
action involved exigent circumstances.
They are repeated here because the
biweekly notice lists all amendments
issued or proposed to be issued
involving no significant hazards
consideration.
For details, see the individual notice
in the Federal Register on the day and
page cited. This notice does not extend
the notice period of the original notice.
sroberts on PROD1PC70 with NOTICES
Florida Power and Light Company,
Docket Nos. 50–250 and 50–251, Turkey
Point Plant, Units 3 and 4, Miami-Dade
County, Florida.
Date of application for amendment:
November 12, 2007.
Brief description of amendment: Use
of alternate method of monitoring rod
position for a control rod or shutdown
rod with an inoperable rod position
indicator.
Date of publication of individual
notice in the Federal Register:
November 28, 2007 (72 FR 67323).
Expiration date of individual notice:
December 28, 2007 (Public comments)
and January 28, 2008 (Hearing requests).
Notice of Issuance of amendments to
Facility Operating Licenses
During the period since publication of
the last biweekly notice, the
Commission has issued the following
amendments. The Commission has
determined for each of these
amendments that the application
complies with the standards and
requirements of the Atomic Energy Act
of 1954, as amended (the Act), and the
Commission’s rules and regulations.
The Commission has made appropriate
findings as required by the Act and the
Commission’s rules and regulations in
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20:08 Dec 28, 2007
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10 CFR Chapter I, which are set forth in
the license amendment.
Notice of Consideration of Issuance of
Amendment to Facility Operating
License, Proposed No Significant
Hazards Consideration Determination,
and Opportunity for A Hearing in
connection with these actions was
published in the Federal Register as
indicated.
Unless otherwise indicated, the
Commission has determined that these
amendments satisfy the criteria for
categorical exclusion in accordance
with 10 CFR 51.22. Therefore, pursuant
to 10 CFR 51.22(b), no environmental
impact statement or environmental
assessment need be prepared for these
amendments. If the Commission has
prepared an environmental assessment
under the special circumstances
provision in 10 CFR 51.22(b) and has
made a determination based on that
assessment, it is so indicated.
For further details with respect to the
action see (1) the applications for
amendment, (2) the amendment, and (3)
the Commission’s related letter, Safety
Evaluation and/or Environmental
Assessment as indicated. All of these
items are available for public inspection
at the Commission’s Public Document
Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville,
Maryland. Publicly available records
will be accessible from the Agencywide
Documents Access and Management
Systems (ADAMS) Public Electronic
Reading Room on the internet at the
NRC Web site, http://www.nrc.gov/
reading-rm/adams.html. If you do not
have access to ADAMS or if there are
problems in accessing the documents
located in ADAMS, contact the PDR
Reference staff at 1 (800) 397–4209,
(301) 415–4737 or by e-mail to
pdr@nrc.gov.
Duke Power Company LLC, Docket Nos.
50–269, 50–270, and 50–287, Oconee
Nuclear Station, Units 1, 2, and 3,
Oconee County, South Carolina.
Date of application of amendments:
January 31, 2007.
Brief description of amendments: The
amendments revised the Technical
Specifications to remove requirements
that are no longer applicable due to the
completion of the control room intake/
booster fan modifications.
Date of Issuance: December 11, 2007.
Effective date: As of the date of
issuance and shall be implemented
within 30 days from the date of
issuance.
Amendment Nos.: 358, 360, and 359.
Renewed Facility Operating License
Nos. DPR–38, DPR–47, and DPR–55:
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Amendments revised the licenses and
the technical specifications.
Date of initial notice in Federal
Register: October 9, 2007 (72 FR 57353)
The Commission’s related evaluation of
the amendments is contained in a Safety
Evaluation dated December 11, 2007.
No significant hazards consideration
comments received: No.
Energy Northwest, Docket No. 50–397,
Columbia Generating Station, Benton
County, Washington.
Date of application for amendment:
July 30, 2007, as supplemented by letter
dated November 6, 2007.
Brief description of amendment: The
changes revise Technical Specification
(TS) 1.4, ‘‘Frequency,’’ TS 3.1.5,
‘‘Control Rod Scram Accumulators,’’ TS
3.4.1, ‘‘Recirculation Loops Operating,’’
TS 3.5.1, ‘‘ECCS [Emergency Core
Cooling System]—Operating,’’ TS 3.5.2,
‘‘ECCS—Shutdown,’’ TS 3.7.1, ‘‘Standby
Service Water (SW) System and
Ultimate Heat Sink (UHS),’’ TS 3.8.1,
‘‘AC [Alternating Current] Sources—
Operating,’’ TS 3.8.2, ‘‘AC Sources—
Shutdown,’’ and TS 5.5.6, ‘‘In-service
Testing Program.’’ The changes include
updates to adopt approved TS Task
Force (TSTF) Travelers 284, Revision 3,
‘‘Add ‘Met’ vs. ‘Perform’ to
Specification 1.4, Frequency,’’ TSTF–
479, Revision 0, ‘‘Changes to Reflect
Revision of 10 CFR 50.55a,’’ and TSTF–
485, Revision 0, ‘‘Correct Example 1.4–
1,’’ and TSTF–497, Revision 0, ‘‘Limit
Inservice Testing Program SR
[Surveillance Requirement] 3.0.2
Application to Frequencies of 2 Years or
Less.’’
Date of issuance: December 13, 2007.
Effective date: As of its date of
issuance and shall be implemented
within 90 days from the date of
issuance.
Amendment No.: 205.
Facility Operating License No. NPF–
21: The amendment revised the Facility
Operating License and Technical
Specifications.
Date of initial notice in Federal
Register: August 28, 2007 (72 FR
49572). The supplement dated
November 6, 2007, provided additional
information that clarified the
application, did not expand the scope of
the application as originally noticed,
and did not change the staff’s original
proposed no significant hazards
consideration determination as initially
published in the Federal Register. The
Commission’s related evaluation of the
amendment is contained in a Safety
Evaluation dated December 13, 2007.
No significant hazards consideration
comments received: No.
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Entergy Gulf States, Inc., and Entergy
Operations, Inc., Docket No. 50–458,
River Bend Station, Unit 1, West
Feliciana Parish, Louisiana.
Date of amendment request: July 2,
2007.
Brief description of amendment: The
amendment modified River Bend
Station, Unit 1, technical specifications
(TSs) requirements for MODE change
limitations in Limiting Condition for
Operation 3.0.4 and Surveillance
Requirement 3.0.4. The TS changes are
consistent with Revision 9 of NRCapproved Industry TS Task Force
(TSTF) Standard TS Change Traveler,
TSTF–359, ‘‘Increase Flexibility in
MODE Restraints.’’ In addition, the
amendment also changed TS Section
1.4, ‘‘Frequency,’’ Example 1.4–1,
‘‘Surveillance Requirements,’’ to
accurately reflect the changes made by
TSTF–359, which is consistent with
NRC-approved TSTF–485, Revision 0,
‘‘Correct Example 1.4–1.’’
Date of issuance: December 6, 2007.
Effective date: As of the date of
issuance and shall be implemented 120
days from the date of issuance.
Amendment No.: 156.
Facility Operating License No. NPF–
47: The amendment revised the Facility
Operating License and Technical
Specifications.
Date of initial notice in Federal
Register: September 11, 2007 (72 FR
51856).
The Commission’s related evaluation
of the amendment is contained in a
Safety Evaluation dated December 6,
2007.
No significant hazards consideration
comments received: No.
sroberts on PROD1PC70 with NOTICES
Indiana Michigan Power Company,
Docket Nos. 50–315, Donald C. Cook
Nuclear Plant, Units 1 and 2 (DCCNP–
1 and DCCNP–2), Berrien County,
Michigan.
Date of application for amendments:
September 15, 2006, as supplemented
on July 25 and October 9, 2007.
Brief description of amendments: The
amendments revised the DCCNP–1 and
DCCNP–2 Technical Specifications (TS)
to allow certain functions in the reactor
protection system and engineered safety
feature actuation system
instrumentation which have installed
bypass test capability to be tested in
bypass. The licensee’s request to correct
the administrative error will be
reviewed and resolved by separate
correspondence.
Date of issuance: December 17, 2007.
Effective date: As of the date of
issuance, and shall be implemented
within 45 days.
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Amendment No.: 300, 283.
Facility Operating License Nos. DPR–
58 and DPR–74: Amendments revise the
License Page and Technical
Specifications.
Date of initial notice in Federal
Register: November 21, 2006 (71 FR
67396)
The supplemental letters contained
clarifying information, did not change
the initial no significant hazards
consideration determination, and did
not expand the scope of the original
Federal Register notice. The
Commission’s related evaluation of the
amendment is contained in a safety
evaluation dated December 17, 2007.
No significant hazards consideration
comments received: No.
Luminant Generation Company LLC,
Docket Nos. 50–445 and 50–446,
Comanche Peak Steam Electric Station,
Unit Nos. 1 and 2, Somervell County,
Texas.
Date of amendment request:
December 19, 2006.
Brief description of amendments: The
amendments revised Technical
Specification 5.5.16, ‘‘Containment
Leakage Rate Testing Program,’’ for
consistency with the requirements of
paragraph 50.55a(g)(4) of Title 10 of the
Code of Federal Regulations for
components classified as Code Class CC.
Date of issuance: December 13, 2007.
Effective date: As of the date of
issuance and shall be implemented
within 120 days from the date of
issuance.
Amendment Nos.: Unit 1–141; Unit
2–141.
Facility Operating License Nos. NPF–
87 and NPF–89: The amendments
revised the Facility Operating Licenses
and Technical Specifications.
Date of initial notice in Federal
Register: May 8, 2007 (72 FR 26179).
The Commission’s related evaluation of
the amendments is contained in a Safety
Evaluation dated December 13, 2007.
No significant hazards consideration
comments received: No.
Nebraska Public Power District, Docket
No. 50–298, Cooper Nuclear Station,
Nemaha County, Nebraska
Date of amendment request: August
16, 2007, as supplemented by letter date
November 5, 2007.
Brief description of amendment: The
amendment revised Technical
Specification 5.5.6, ‘‘Inservice Testing
Program,’’ to allow a one-time extension
of the 5-year frequency requirement for
setpoint testing of safety valve MS–RV–
70ARV.
Date of issuance: December 4, 2007.
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Effective date: As of the date of
issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 228.
Facility Operating License No. DPR–
46: Amendment revised the Facility
Operating License and Technical
Specifications.
Date of initial notice in Federal
Register: September 25, 2007 (72 FR
54476). The supplement dated
November 5, 2007, provided additional
information that clarified the
application, did not expand the scope of
the application as originally noticed,
and did not change the staff’s original
proposed no significant hazards
consideration determination as initially
published in the Federal Register. The
Commission’s related evaluation of the
amendment is contained in a Safety
Evaluation dated December 4, 2007.
No significant hazards consideration
comments received: No.
STP Nuclear Operating Company,
Docket Nos. 50–498 and 50–499, South
Texas Project, Units 1 and 2, Matagorda
County, Texas
Date of amendment request: February
28, 2007, as supplemented by letter
dated May 22, 2007.
Brief description of amendments: The
amendments revise the language in the
Technical Specifications to conform to
the licensing basis as established by
Amendment Nos. 87 and 74, for Units
1 and 2, respectively, dated May 27,
1997.
Date of issuance: December 6, 2007.
Effective date: As of the date of
issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: Unit 1–181; Unit
2–168.
Facility Operating License Nos. NPF–
76 and NPF–80: The amendments
revised the Facility Operating Licenses
and Technical Specifications.
Date of initial notice in Federal
Register: May 22, 2007 (72 FR 28723).
The supplement dated May 22, 2007,
provided additional information that
clarified the application, did not expand
the scope of the application as originally
noticed, and did not change the staff’s
original proposed no significant hazards
consideration determination as
published in the Federal Register. The
Commission’s related evaluation of the
amendments is contained in a Safety
Evaluation dated December 6, 2007.
No significant hazards consideration
comments received: No.
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Nuclear Management Company, LLC,
Docket Nos. 50–282 and 50–306, Prairie
Island Nuclear Generating Plant
(PINGP), Units 1 and 2, Goodhue
County, Minnesota.
Date of application for amendments:
December 14, 2006, supplemented by
letter dated November 13, 2007.
Brief description of amendments: The
amendments revise the sump debris
interceptor nomenclature in PINGP Unit
1 and Unit 2 Technical Specifications
(TS) 3.5.2 to more clearly reflect the
configuration of the new Emergency
Core Cooling System sump strainers that
were installed to address Generic Letter
2004–02, ‘‘Potential Impact of Debris
Blockage on Emergency Recirculation
During Design Basis Accidents at
Pressurized-Water Reactors.’’ The
amendments also revise the required
Refueling Water Storage Tank (RWST)
water level in TS 3.5.4 to reflect the
administratively controlled water
inventory in the RWST.
Date of issuance: December 14, 2007.
Effective date: As of the date of
issuance and shall be implemented
within 90 days.
Amendment Nos.: 182/172.
Facility Operating License Nos. DPR–
42 and DPR–60: Amendments revised
the Facility Operations License and
Technical Specifications.
Date of initial notice in Federal
Register: February 27, 2007 (72 FR
8804)
The supplemental letter contained
clarifying information and did not
change the initial no significant hazards
consideration determination, and did
not expand the scope of the original
Federal Register notice. The
Commission’s related evaluation of the
amendments is contained in a Safety
Evaluation dated December 14, 2007.
No significant hazards consideration
comments received: No.
sroberts on PROD1PC70 with NOTICES
Omaha Public Power District, Docket
No. 50–285, Fort Calhoun Station, Unit
No. 1, Washington County, Nebraska.
Date of amendment request: July 31,
2007.
Brief description of amendment: The
amendment revises Technical
Specification (TS) 2.7(1), (Electrical
Systems—Minimum Requirements,’’ TS
2.7(2), (‘‘Electrical Systems—
Modification of Minimum
Requirements,’’ and TS 3.7(5),
‘‘Emergency Power System Periodic
Tests—Required Safety Related
Inverters.’’ The licensee is adding two
safety-related swing inverters to the 120
Volt alternating current instrument
buses. The TS changes reflect
modifications made to the plant and are
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needed to take advantage of the
additional operational flexibility the
swing inverters will provide.
Date of issuance: December 17, 2007.
Effective date: As of its date of
issuance and shall be implemented
within 90 days from the date of
issuance.
Amendment No.: 251.
Renewed Facility Operating License
No. DPR–40: The amendment revised
the Technical Specifications.
Date of initial notice in Federal
Register: August 28, 2007 (72 FR
49582). The Commission’s related
evaluation of the amendment is
contained in a safety evaluation dated
December 17, 2007.
No significant hazards consideration
comments received: No.
2. The NRC seeks to obtain
information on licensee administrative
and managerial controls to deter and
address inattentiveness and complicity
among licensee security personnel
including contractors and
subcontractors.
3. This bulletin requires that
addressees provide a written response to
the NRC in accordance with Title 10 of
the Code of Federal Regulations (10
CFR), Section 50.54(f) or 10 CFR
70.22(d).
This Federal Register notice is
available through the NRC’s
Agencywide Documents Access and
Management System (ADAMS) under
accession number ML073480342.
Dated at Rockville, Maryland, this 21st day
of December, 2007.
For The Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor
Licensing, Office of Nuclear Reactor
Regulation.
[FR Doc. E7–25416 Filed 12–28–07; 8:45 am]
DATES:
BILLING CODE 7590–01–P
Security Officer Attentiveness
Nuclear Regulatory
Commission.
ACTION: Notice of issuance.
AGENCY:
SUMMARY: All holders of operating
licenses for nuclear power reactors,
except those who have permanently
ceased operation and have certified that
fuel has been removed from the reactor
vessel, and Category I fuel facilities. The
contents of this bulletin are for
information to Category III fuel
facilities, independent spent fuel storage
installations, conversion facilities and
gaseous diffusion plants. The U.S.
Nuclear Regulatory Commission (NRC)
is issuing this bulletin to achieve the
following three objectives:
1. The agency is notifying addressees
about the NRC staff’s need for
information associated with licensee
security program administrative and
management controls as a result of
security personnel inattentiveness,
especially involving complicity, and
related concerns with the behavior
observation program (BOP). The
information is needed to determine if
further regulatory action is warranted, if
the necessary inspection program needs
to be enhanced, or if additional
assessment of security program
implementation is needed.
Frm 00100
Fmt 4703
Sfmt 4703
ADDRESSES:
Not applicable.
FOR FURTHER INFORMATION CONTACT:
Timothy S. McCune at 301–415–6474 or
by email tsm5@nrc.gov, Kevin Ramsey
at 301–415–3123 or by e-mail
kmr@nrc.gov, or Merrilee Banic at 301–
415–2771 or email mjb@nrc.gov.
NRC
Bulletin 2007–01 may be examined,
and/or copied for a fee, at the NRC’s
Public Document Room at One White
Flint North, 11555 Rockville Pike (first
floor), Rockville, Maryland. Publicly
available records will be accessible
electronically from the Agencywide
Documents Access and Management
System (ADAMS) Public Electronic
Reading Room on the Internet at the
NRC Web site, http://www.nrc.gov/NRC/
ADAMS/index.html. The ADAMS
number for the bulletin is
ML051740058.
If you do not have access to ADAMS
or if you have problems in accessing the
documents in ADAMS, contact the NRC
Public Document Room (PDR) reference
staff at 1–800–397–4209 or 301–415–
4737 or by e-mail to pdr@nrc.gov.
SUPPLEMENTARY INFORMATION:
NUCLEAR REGULATORY
COMMISSION
PO 00000
The bulletin was issued on
December 12, 2007.
Dated at Rockville, Maryland, this 19th day
of December 2007.
For the Nuclear Regulatory Commission.
Martin C. Murphy,
Chief, Generic Communications Branch,
Division of Policy and Rulemaking, Office
of Nuclear Reactor Regulation.
[FR Doc. E7–25398 Filed 12–28–07; 8:45 am]
BILLING CODE 7590–01–P
E:\FR\FM\31DEN1.SGM
31DEN1

Agencies

[Federal Register Volume 72, Number 249 (Monday, December 31, 2007)]
[Notices]
[Pages 74354-74365]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-25416]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from December 6, 2007 to December 19, 2007. The
last biweekly notice was published on December 18, 2007 (72 FR 71703).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed no Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received
[[Page 74355]]
within 30 days after the date of publication of this notice will be
considered in making any final determination. Within 60 days after the
date of publication of this notice, the licensee may file a request for
a hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, Division of Administrative Services,
Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the Commission's
Public Document Room (PDR), located at One White Flint North, Public
File Area O1F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
Within 60 days after the date of publication of this notice,
person(s) may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
via electronic submission through the NRC E-Filing system for a hearing
and a petition for leave to intervene. Requests for a hearing and a
petition for leave to intervene shall be filed in accordance with the
Commission's ``Rules of Practice for Domestic Licensing Proceedings''
in 10 CFR part 2. Interested person(s) should consult a current copy of
10 CFR 2.309, which is available at the Commission's PDR, located at
One White Flint North, Public File Area 01F21, 11555 Rockville Pike
(first floor), Rockville, Maryland. Publicly available records will be
accessible from the Agencywide Documents Access and Management System's
(ADAMS) Public Electronic Reading Room on the Internet at the NRC Web
site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request
for a hearing or petition for leave to intervene is filed within 60
days, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for hearing or a petition for leave to intervene must be
filed in accordance with the NRC E-Filing rule, which the NRC
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process
requires participants to submit and serve documents over the Internet
or in some cases to mail copies on electronic storage media.
Participants may not submit paper copies of their filings unless they
seek a waiver in accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
five (5) days prior to the filing deadline, the petitioner/requestor
must contact the Office of the Secretary by e-mail at
HEARINGDOCKET@NRC.GOV, or by
[[Page 74356]]
calling (301) 415-1677, to request (1) a digital ID certificate, which
allows the participant (or its counsel or representative) to digitally
sign documents and access the E-Submittal server for any proceeding in
which it is participating; and/or (2) creation of an electronic docket
for the proceeding (even in instances in which the petitioner/requestor
(or its counsel or representative) already holds an NRC-issued digital
ID certificate). Each petitioner/requestor will need to download the
Workplace Forms Viewer\TM\ to access the Electronic Information
Exchange (EIE), a component of the E-Filing system. The Workplace Forms
Viewer\TM\ is free and is available at http://www.nrc.gov/site-help/e-
submittals/install-viewer.html. Information about applying for a
digital ID certificate is available on NRC's public Web site at http://
www.nrc.gov/site-help/e-submittals/apply-certificates.html.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it can then submit
a request for hearing or petition for leave to intervene. Submissions
should be in Portable Document Format (PDF) in accordance with NRC
guidance available on the NRC public Web site at http://www.nrc.gov/
site-help/e-submittals.html. A filing is considered complete at the
time the filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE system no later than
11:59 p.m. Eastern Time on the due date. Upon receipt of a
transmission, the E-Filing system time-stamps the document and sends
the submitted an e-mail notice confirming receipt of the document. The
EIE system also distributes an e-mail notice that provides access to
the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically may seek assistance through the
``Contact Us'' link located on the NRC Web site at http://www.nrc.gov/
site-help/e-submittals.html or by calling the NRC technical help line,
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time,
Monday through Friday. The help line number is (800) 397-4209 or
locally, (301) 415-4737.
Participants who believe that they have a good cause for not
submitting documents electronically must file a motion, in accordance
with 10 CFR 2.302(g), with their initial paper filing requesting
authorization to continue to submit documents in paper format. Such
filings must be submitted by: (1) First class mail addressed to the
Office of the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff; or (2) courier, express mail, or expedited
delivery service to the Office of the Secretary, Sixteenth Floor, One
White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852,
Attention: Rulemaking and Adjudications Staff. Participants filing a
document in this manner are responsible for serving the document on all
other participants. Filing is considered complete by first-class mail
as of the time of deposit in the mail, or by courier, express mail, or
expedited delivery service upon depositing the document with the
provider of the service.
Non-timely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission, the presiding
officer, or the Atomic Safety and Licensing Board that the petition
and/or request should be granted and/or the contentions should be
admitted, based on a balancing of the factors specified in 10 CFR
2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later
than 11:59 p.m. Eastern Time on the due date.
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant
to an order of the Commission, an Atomic Safety and Licensing Board, or
a Presiding Officer. Participants are requested not to include personal
privacy information, such as social security numbers, home addresses,
or home phone numbers in their filings. With respect to copyrighted
works, except for limited excerpts that serve the purpose of the
adjudicatory filings and would constitute a Fair Use application,
participants are requested not to include copyrighted materials in
their submission.
For further details with respect to this amendment action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to pdr@nrc.gov.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina.
Date of amendment request: November 19, 2007.
Description of amendment request: The proposed amendment would make
administrative revisions to delete requirements that are obsolete or
redundant, or correct and clarify the typing and formatting of other
requirements. The proposed changes will not result in changes to the
plant design or the procedural controls for the operation,
surveillance, or maintenance of the plant.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the Proposed Changes Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated?
No. The proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed changes are administrative. The changes
delete obsolete or redundant requirements, clarify existing
requirements, and correct typing and formatting errors. There will
be no resulting changes to the plant design or procedural controls.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the Proposed Changes Create the Possibility of a New or
Different Kind of Accident From Any Previously Evaluated?
No. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated. There
are no physical changes being made to the plant or to the manner in
which the plant is operated. Therefore, the changes do not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Do the Proposed Changes Involve a Significant Reduction in
the Margin of Safety?
No. The proposed changes do not involve a significant reduction
in the margin of safety. There are no physical changes being made to
the plant or to the manner in which
[[Page 74357]]
the plant is operated. The proposed changes are administrative. The
changes delete obsolete or redundant requirements, clarify existing
requirements, and correct typing and formatting errors. Therefore,
the changes do not involve a significant reduction in any margin of
safety for HBRSEP [H.B. Robinson Steam Electric Plant], Unit No. 2.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal. Department, Progress Energy Service Company, LLC, Post
Office Box 1551, Raleigh, North Carolina 27602.
NRC Branch Chief: Thomas H. Boyce.
Duke Power Company LLC, Docket No. 50-369, McGuire Nuclear Station,
Unit 1, Mecklenburg County, North Carolina.
Date of amendment request: February 21, 2007, as supplemented
August 9, 2007.
Description of amendment request: The proposed amendment would
allow, on a one-time basis, an extension of the interval governing the
conduct of the Integrated Leak Rate Test (ILRT) for McGuire Nuclear
Station, Unit 1. The proposed amendment would revise administrative
Technical Specification (TS) 5.5.2, ``Containment Leak Rate Testing
Program,'' from the currently approved 15-year interval (since the last
McGuire Nuclear Station, Unit 1, Type A test) to a frequency
encompassing the end of the McGuire Nuclear Station, Unit 1, End-of-
Cycle 19 refueling outage (approximately 6 months beyond the present TS
frequency).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated, or
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated, or
3. Involve a significant reduction in a margin of safety.
First Standard
The proposed amendment will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed extension to the Type A testing intervals
cannot increase the probability of an accident previously evaluated
since extension of the intervals is not a physical plant
modification that could alter the probability of accident
occurrence, nor is it an activity or modification by itself that
could lead to equipment failure or accident initiation. The proposed
extension to the Type A testing intervals does not result in a
significant increase in the consequences of an accident as
documented in NUREG-1493 [``Performance-Based Containment Leak-Test
Program'', NUREG-1493, September 1995]. The NUREG notes that very
few potential containment leakage paths are not identified by Type B
and Type C tests. It concludes that reducing the Type A testing
frequency to once per twenty years leads to an imperceptible
increase in risk. McGuire [Nuclear Station, Unit 1 (McGuire Unit 1)]
provides a high degree of assurance through testing and inspection
that the containment will not degrade in a manner detectable only by
Type A testing. Prior Type A tests for McGuire Unit 1 identified
containment leakage within acceptance criteria, indicating a very
leak tight containment. Inspections required by the ASME Code
[American Society of Mechanical Engineers (ASME), Boiler and
Pressure Vessel Code (Code)] are also performed in order to identify
indications of containment degradation that could affect leak
tightness. Separately, Type B and Type C testing, required by TS
[Technical Specification] identify any containment opening from
design penetrations, such as valves, that would otherwise be
detected by a Type A test. These factors establish that an extension
to the Type A test intervals will not represent a significant
increase in the consequences of an accident.
Second Standard
The proposed amendments will not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The proposed revisions to the McGuire TS add a one-time
extension to the current interval for Type A testing. The current
test interval of fifteen years, based on past performance, would be
extended on a one-time basis to approximately fifteen and a half
years from the last Type A test. The proposed extension to the Type
A test interval does not create the possibility of a new or
different type of accident since there are no physical changes being
made to the plants and there are no changes to the operation of the
plants that could introduce a new failure mode.
Third Standard
The proposed amendment will not involve a significant reduction
in a margin of safety. The proposed revisions to the McGuire TS add
a one-time extension to the current interval for Type A testing. The
current test interval of fifteen years, based on past performance,
would be extended on a one-time basis to approximately fifteen and a
half years from the last Type A test. The proposed extension to Type
A test intervals will not significantly reduce the margin of safety.
The NUREG-1493 generic study of the effects of extending containment
leakage testing intervals found that a twenty-year interval resulted
in an imperceptible increase in risk to the public. NUREG-1493 found
that, generically, the design containment leakage rate contributes
about 0.1 percent of the overall risk and that decreasing the Type A
testing frequency would have a minimal effect on this risk, since 95
percent of the Type A detectable leakage paths would already be
detected by Type B and Type C testing. Similar proposed changes have
been previously reviewed and approved by the NRC, and they are
applicable to McGuire. Based upon the preceding discussion, Duke
Energy Corporation [Duke Power Company, LLC] has concluded that the
proposed amendments do not involve a significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Evangelos C. Marinos.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana.
Date of amendment request: November 15, 2007.
Description of amendment request: The proposed change would
relocate Surveillance Requirement (SR) 3.8.3.6 from the Technical
Specifications (TS) to a licensee-controlled document. SR 3.8.3.6
requires the Emergency Diesel Generator (EDG) Fuel Oil Storage Tanks
(FOSTs) to be drained, sediment removed, and cleaned on a 10-year
interval.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The FOSTs provide the storage for the EDG fuel oil, assuring an
adequate volume is available for each EDG to operate for seven days
in the event of a loss of offsite power concurrent with a loss of
coolant accident. The relocation of the SR to drain and clean the
FOSTs will not impact any of the previously analyzed accidents.
Sediment in the tank, or failure to perform this SR, does not
necessarily result in an inoperable storage tank. Fuel oil quantity
and quality are assured by other TS SRs which remain unchanged.
These SRs help ensure tank sediment is minimized and ensure that any
[[Page 74358]]
degradation of the tank wall surface that results in a fuel oil
volume reduction is detected and corrected in a timely manner. As a
result, adequate controls exist to allow relocation of this
preventative maintenance cleaning requirement to licensee controlled
documents.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed TS changes do not involve the addition or
modification of any plant equipment. Also, the proposed change will
not alter the design configuration, or method of operation of plant
equipment beyond its normal functional capabilities. The proposed TS
change does not create any new credible failure mechanisms,
malfunctions or accident initiators.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not alter or exceed a design basis or
safety limit. Diesel generator fuel oil quantity and quality will
continue to be maintained within acceptable limits of the TS to
assure the ability of the EDG to perform its intended function.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hilt.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi.
Date of amendment request: December 5, 2007.
Description of amendment request: The proposed amendment would
change the Grand Gulf Nuclear Station, Unit 1 (GGNS), Technical
Specification (TS) 5.6.5, ``Core Operating Limits Report (COLR),'' to
add a reference to an analytical method that will be used to determine
core operating limits. The new reference, NEDC-33383P, ``GEXL97
Correlation Applicable to ATRIUM-10 Fuel,'' will allow Entergy
Operations, Inc. (Entergy) to use a Global Nuclear Fuel (GNF) method to
determine fuel assembly critical power of AREVA ATRIUM-10 fuel. GGNS
currently operates with a full core of ATRIUM-10 fuel. Entergy plans to
use the GEXL97 correlation for GGNS operating Cycle 17 currently
scheduled to begin in the fall 2008. Additionally, an administrative
change is proposed to an existing reference in TS 5.6.5.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Core operating limits are established each operating cycle in
accordance with TS 3.2, ``Power Distribution'' and TS 5.6.5, ``Core
Operating Limits Report (COLR)''. These core operating limits ensure
that the fuel design limits are not exceeded during any conditions
of normal operation or in the event of any Anticipated Operational
Occurrence (AOO). The methods used to determine the operating limits
are those previously found acceptable by the NRC and listed in TS
Section 5.6.5.b.
A change to TS 5.6.5.b is requested to include an additional
reference to the list of analytical methods. GGNS currently operates
with a full core of AREVA ATRIUM-10 fuel but is scheduled to load
GE14 fuel during the next refueling outage. GGNS plans to use the
analysis methods of the new fuel vendor, GNF for the analysis of the
mixed core. The GEXL97 correlation accurately models predicted core
behavior and appropriately determines the overall critical power
uncertainty of the method. In addition, the GEXL97 application range
covers the range of expected operation of the ATRIUM-10 fuel during
normal steady state and transient conditions in the GGNS reload
cores. Although a depressurization transient could result in vessel
pressures below the range of GEXL97, the transient would not
threaten fuel cladding integrity, since the margin to the MCPR
[minimum critical power ratio] safety limit increases with
decreasing reactor pressure.
Additionally, Entergy proposes an administrative change to the
GESTAR-Il reference in TS 5.6.5.b. The administrative change does
not alter any method of analysis as described in the NRC approved
versions of GESTAR-II. The requested TS changes concern the use of
analytical methods and do not involve any plant modifications or
operational changes that could affect any postulated accident
precursors or accident mitigation systems and do not introduce any
new accident initiation mechanisms. The proposed changes have no
effect on the type or amount of radiation released, and have no
effect on predicted offsite doses in the event of an accident. Thus,
the proposed change does not affect the probability of an accident
previously evaluated nor does it increase the radiological
consequences of any accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed TS changes will not change the design function,
reliability, performance, or operation of any plant systems,
components, or structures. It does not create the possibility of a
new failure mechanism, malfunction, or accident initiators not
considered in the design and licensing bases. Plant operation will
continue to be within the core operating limits that are established
using NRC approved methods that are applicable to the GGNS design
and the GGNS fuel.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change adds GEXL97 to the list of analytical
methods in TS 5.6.5.b that can be used to determine core operating
limits. Use of the GEXL97 correlation analytical method provides an
equivalent level of protection as that currently provided. The
administrative change does not alter any method of analysis as
described in the NRC approved versions of GESTAR-II. The proposed
change does not modify the safety limits or set points at which
protective actions are initiated, and does not change the
requirements governing operation or availability of safety equipment
assumed to operate to preserve the margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hiltz.
[[Page 74359]]
Exelon Generation Company, LLC, Docket No. 50-353, Limerick Generating
Station, Unit 2, Montgomery County, Pennsylvania.
Date of amendment request: November 16, 2007.
Description of amendment request: The proposed changes revise
technical specification (TS) action requirements associated with
inoperable reactor coolant system (RCS) leakage detection systems. A
new TS action requirement is proposed that will address the
inoperability of the drywell unit cooler condensate flow rate
monitoring system concurrent with one other RCS leakage detection
system, other than the primary containment atmosphere gaseous
radioactivity monitoring system. This would relax the allowed out-of-
service time for the specified combination of systems and is related to
the current inoperability of the drywell unit cooler condensate flow
rate monitoring system. The proposed changes would be effective for the
remainder of the current operating cycle (Cycle 10), which is currently
scheduled to end in the spring of 2009, or until the next shutdown of
sufficient duration to allow for drywell unit cooler condensate flow
rate monitoring system repairs, whichever comes first.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes continue to maintain an acceptable level of
reactor coolant system (RCS) leakage detection instrumentation
required to support plant operations. The level of RCS leakage
detection capability inherent with the proposed changes will
continue to provide acceptable early warning detection of potential
RCS pressure boundary degradation. The proposed changes do not
impact the physical configuration or design function of plant
structures, systems, or components (SSCs) or the manner in which
SSCs are operated, modified, tested, or inspected [with the
exception of an increase in allowed out-of-service time for a
concurrent inoperability of the drywell unit cooler condensate flow
rate monitoring system and another specified RCS leakage detection
system]. The proposed changes do not impact the initiators or
assumptions of analyzed events, nor do they impact mitigation of
accidents or transient events. Therefore, the proposed changes do
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes only affect systems associated with the
detection of leakage resulting from the degradation of the RCS
pressure boundary. The proposed changes do not alter plant
configuration or require that new plant equipment be installed. The
RCS leakage detection systems will continue to function as designed
in all modes of operation. No new accident type is created as a
result of the proposed changes. No new failure mode for any
equipment is created. The proposed changes do not alter assumptions
made about accidents previously evaluated. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not involve any physical changes to
plant SSCs or the manner in which SSCs are operated, modified,
tested, or inspected. The proposed changes do not involve a change
to any safety limits, limiting safety system settings, limiting
conditions of operation, or design parameters for any SSC. The
proposed changes do not impact any safety analysis assumptions and
do not involve a change in initial conditions, system response
times, or other parameters affecting an accident analysis.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, with changes as noted above, it appears that the three
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: J. Bradley Fewell, Esquire, Associate
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois.
Date of amendment request: October 9, 2007.
Description of amendment request: A change is proposed to the
technical specifications (TS) of Quad Cities Nuclear Power Station
(QCNPS), Units 1 and 2, consistent with Technical Specifications Task
Force (TSTF) Change Traveler TSTF-423 to the standard TSs for boiling
water reactor plants, to allow, for some systems, entry into hot
shutdown rather than cold shutdown to repair equipment, if risk is
assessed and managed consistent with the program in place for complying
with the requirements of Title 10 of the Code of Federal Regulations
(10 CFR) Section 50.65(a)(4). Changes proposed herein will be made to
the QCNPS, Units 1 and 2, TSs for selected required action end states
providing this allowance.
The licensee reviewed the proposed no significant hazards
consideration (NSHC) determination published in the Federal Register on
March 23, 2007 (71 FR 14726) and concluded that it is applicable to
QCNPS, Units 1 and 2. The licensee incorporated the proposed
determination by reference to satisfy the requirements of 10 CFR
50.91(a).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC is presented below:
1. The Proposed Change Does Not Involve a Significant Increase
in the Probability or Consequences of an Accident Previously
Evaluated.
The proposed change allows a change to certain required end
states when the TS Completion Times for remaining in power operation
will be exceeded. Most of the requested technical specification (TS)
changes are to permit an end state of hot shutdown (Mode 3) rather
than an end state of cold shutdown (Mode 4) contained in the current
TS. The request was limited to: (1) Those end states where entry
into the shutdown mode is for a short interval, (2) entry is
initiated by inoperability of a single train of equipment or a
restriction on a plant operational parameter, unless otherwise
stated in the applicable technical specification, and (3) the
primary purpose is to correct the initiating condition and return to
power operation as soon as is practical. Risk insights from both the
qualitative and quantitative risk assessments were used in specific
TS assessments. Such assessments are documented in Section 6 of GE
NEDC-32988, Revision 2, ``Technical Justification to Support Risk
Informed Modification to Selected Required Action End States for BWR
Plants.'' They provide an integrated discussion of deterministic and
probabilistic issues, focusing on specific technical specifications,
which are used to support the proposed TS end state and associated
restrictions. The staff finds that the risk insights support the
conclusions of the specific TS assessments. Therefore, the
probability of an accident previously evaluated is not significantly
increased, if at all. The consequences of an accident after adopting
proposed TSTF-423, are no different than the consequences of an
accident prior to adopting TSTF-423. Therefore, the consequences of
an accident previously evaluated are not significantly affected by
this change. The addition of a requirement to assess and manage the
risk introduced by this change will further minimize possible
concerns. Therefore, this
[[Page 74360]]
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The Proposed Change Does Not Create the Possibility of a New
or Different Kind of Accident From Any Previously Evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
If risk is assessed and managed, allowing a change to certain
required end states when the TS Completion Times for remaining in
power operation are exceeded, i.e., entry into hot shutdown rather
than cold shutdown to repair equipment, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change and the commitment by the licensee to adhere to the guidance
in TSTF-IG-05-02, Implementation Guidance for TSTF-423, Revision 0,
``Technical Specifications End States, NEDC-32988-A,'' will further
minimize possible concerns. Thus, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
3. The Proposed Change Does Not Involve a Significant Reduction
in the Margin of Safety.
The proposed change allows, for some systems, entry into hot
shutdown rather than cold shutdown to repair equipment, if risk is
assessed and managed. The BWROG's [Boiling Water Reactor Owners
Group's] risk assessment approach is comprehensive and follows staff
guidance as documented in RGs [Regulatory Guides] 1.174 and 1.177.
In addition, the analyses show that the criteria of the three-tiered
approach for allowing TS changes are met. The risk impact of the
proposed TS changes was assessed following the three-tiered approach
recommended in RG 1.177. A risk assessment was performed to justify
the proposed TS changes. The net change to the margin of safety is
insignificant. Therefore, this change does not involve a significant
reduction in a margin of safety.
Therefore, the NRC staff proposes to determine that the requested
amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
Luminant Generation Company LLC, Docket Nos. 50-445 and 50-446,
Comanche Peak Steam Electric Station, Units 1 and 2, Somervell County,
Texas
Date of amendment request: November 29, 2007.
Brief description of amendments: Revision to Technical
Specification (TS) 3.6.7, (``Spray Additive System,'' to allow
modifications to the facility potentially required to comply with U.S.
Nuclear Regulatory Commission (NRC) Generic Letter 2004-02, ``Potential
Impact of Debris Blockage on Emergency Recirculation during Design
Basis Accident at Pressurized Water Reactors.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No
The proposed change[s] [do] not impact the initiation or
probability of occurrence of any accident.
The accidents evaluated in the Final Safety Analysis report
(FSAR) that could be affected by this proposed change are those
involving the pressurization of the containment and those involving
recirculation of fluid within the Emergency Core Cooling System
(ECCS) or the Containment Spray System (e.g., loss of coolant
accidents (LOCAs)).
The change to a minimum pH [potential of Hydrogen] of 7.1 will
not result in a significant increase in the radiological
consequences of a LOCA as-described below.
The equilibrium spray pH during the recirculation phase
resulting from this change will be greater than or equal to 7.1. The
pH range for the spray will be bounded by the water spray solution
which is borated water with a maximum of 2600 parts per million
(ppm) boron buffered to a final spray solution pH much less than the
10.5 as described in the current FSAR Section 3.11(B) for the
postulated spray solution environment. The maximum pH is the
limiting parameter for equipment qualification. Since the resulting
pH level will be closer to neutral using the lower limit of 7.1,
post-LOCA corrosion of containment components will not be increased.
Post-LOCA hydrogen generation will be reduced. There will not be an
adverse radiation dose effect on any safety-related equipment. Thus,
the potential for failures of the ECCS or safety-related equipment
following a LOCA will not be increased as a result of the proposed
change.
This modification affects the Containment Spray System which is
intended to respond to and mitigate the effects of a LOCA. The
Containment Spray System will continue to function in a manner
consistent with the plant design basis. There will be no degradation
in the performance of nor an increase in the number of challenges to
equipment assumed to function during an accident situation.
Therefore, these Technical Specification (TS) revisions do not
affect the probability of any event initiators. There will be no
adverse changes to normal plant operating parameters, Engineered
Safety Features (ESF) actuation setpoints, or accident mitigation
capabilities.
The proposed change allows the Spray Additive System currently
used to mitigate the consequences of an accident to maintain the
equilibrium sump pH at greater than or equal to 7.1 to minimize
chloride-induced stress corrosion cracking in austenitic stainless
components important to safety located inside containment.
Therefore, the proposed changes will not increase the probability of
an accident or malfunction of equipment important to safety
previously evaluated in the FSAR.
The offsite and control room doses will continue to meet the
requirements of [Title 10 of the Code of Federal Regulations (10
CFR) part 100] 10 CFR 100, 10 CFR 50 Appendix A [General Design
Criterion] GDC 19, [Standard Review Plan] SRP 15.6.5.11, and SRP
6.4.11. The proposed new pH limit will provide satisfactory
retention of iodine in the sump water, as well as provide adequate
pH control to minimize the potential of chloride-induced stress
corrosion cracking of austenitic stainless steel components.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change to the revised Surveillance for the
Containment Spray Additive System provides for a required minimum
equilibrium pH in containment post accident. There are no electrical
or mechanical components being added whose failure could prevent the
system from functioning.
No new accident scenarios, transient precursors, or limiting
single failures are introduced as a result of the proposed changes.
There will be no adverse effect or challenges imposed on any safety-
related system as a result of this proposed change. The amount of
sodium hydroxide (NaOH) will provide a minimum equilibrium sump pH
of 7.1 following mixing. Therefore, the possibility of a new or
different type of accident is not created.
There are no changes which would cause the malfunction of
safety-related equipment, assumed to be operable in the accident
analyses, as a result of the proposed Technical Specification
changes. The possibility of a malfunction of safety-related
equipment with a different result is not created.
Therefore, the proposed change[s] [do] not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No
The only function of the chemical additive system is to provide
pH control of the post-accident containment recirculation sump
water, since the borated water from the Refueling Water Storage Tank
(RWST) used as the containment spray pump suction source during
injection is sufficient to remove iodine from the containment
atmosphere following a LOCA. The net effect on the pH control
function of reducing the
[[Page 74361]]
amount of buffer is that the equilibrium sump pH will be lowered to
a minimum of 7.1. There will be no change to the current Technical
Specification acceptance limits on RWST volume and boron
concentration. The resulting equilibrium sump pH level from this
change will be closer to neutral; therefore, the post-LOCA corrosion
of containment components will not be increased (i.e., would be
reduced).
Because the long term pH will be maintained greater than or
equal to 7.1, margin to minimize the potential for stress corrosion
cracking is maintained.
The radiological analysis, as discussed in the technical
analysis above, is shown not to be impacted. There will be no change
to the [departure from nucleate boiling ratio] DNBR Correlation
Limit, the design DNBR limits, or the safety analysis DNBR limits
discussed in Bases Section 2.1.1. There will be no effect on the
manner in which Safety Limits or Limiting Safety System Settings are
determined nor will there be any effect on those plant systems
necessary to assure the accomplishment of protection functions.
There will be no adverse impact on Departure of Nucleate Boiling
Ratio limits, [heat flux hot channel factor] FQ, [nuclear
enthalpy rise hot channel factor] F-delta-H, LOCA peak cladding
temperature, peak local power density, or any other margin of
safety.
Therefore the proposed change[s] [do] not involve a reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Timothy P. Matthews, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: Thomas G. Hiltz.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California.
Date of amendment requests: October 2, 2007.
Description of amendment requests: The proposed amendments would
revise Technical Specification (TS) 3.5.4, ``Refueling Water Storage
Tank (RWST),'' Surveillance Requirement (SR) 3.5.4.2, to increase the
minimum required borated water volume from ``>= [greater than or equal
to] 400,000 gallons (81.5% indicated level)'' to ``>= 455,300 gallons
(93.6% level),'' to reflect the new sump design required to comply with
U.S. Nuclear Regulatory Commission (NRC) Generic Letter 2004-02,
``Potential Impact of Debris Blockage on Emergency Recirculation during
Design-Basis Accident at Pressurized Water Reactors.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. [Do] the proposed change[s] involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change[s] [revise] the minimum RWST borated water
volume. The RWST borated water volume is not an initiator of any
accident previously evaluated. As a result, the probability of an
accident previously evaluated is not affected. The proposed
change[s] [do] not alter or prevent the ability of structures,
systems, and components from performing their intended function to
mitigate the consequences of an initiating event within the assumed
acceptance limits. The effect on containment flood level, equipment
qualification, and containment sump pH remain within the limits
assumed in the design and accident analyses. The proposed change[s]
[do] not affect the source term, containment isolation, or
radiological release assumptions used in evaluating the radiological
consequences of an accident previously evaluated. Further, the
proposed change[s] [do] not increase the types or amounts of
radioactive effluent that may be released offsite, nor significantly
increase individual or cumulative occupational/public radiation
exposures. The proposed change[s] are consistent with the safety
analysis assumptions and resultant consequences.
Therefore, the proposed change[s] [do] not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. [Do] the proposed change[s] create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The change[s] [do] not involve a physical alteration of the
plant (i.e., no new or different components or physical changes are
involved with this change) or a change in the methods governing
normal plant operation. The change[s] [do] not alter any assumptions
made in the safety analysis.
Therefore, the proposed change[s] will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. [Do] the proposed change[s] involve a significant reduction
in a margin of safety?
Response: No.
The proposed change[s] to revise the required RWST minimum
borated water volume [do] not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not affected by [these] change[s]. The proposed change[s] will
not result in plant operation in a configuration outside of the
design basis.
Therefore, the proposed change[s] [do] not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c)
are satisfied. Therefore, the NRC staff proposes to determine that
the amendment requests involve no significant hazards consideration.
Attorney for licensee: Jennifer Post, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Branch Chief: Thomas G. Hiltz.
Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power
Plant (HBPP), Unit 3 Humboldt County, California.
Date of amendment request: November 5, 2007.
Description of amendment request: The licensee has proposed
amending the technical specifications (TS) to delete many
operational and administrative requirements upon transfer of spent
nuclear fuel assemblies and fuel fragment containers from the Spent
Fuel Pool (SFP) to the Humboldt Bay Independent Spent Fuel Storage
Installation (ISFSI). Some TS requirements will be relocated to the
HBPP Quality Assurance Plan.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes reflect the transfer of spent fuel from the
Spent Fuel Pool to the Humboldt Bay (HB) Independent Spent Fuel
Storage Installation. Design basis accidents related to the SFP are
discussed in the Humboldt Bay Power Plant Unit 3 Defueled Safety
Analysis Report (DSAR). These postulated accidents are predicated on
spent fuel being stored in the SFP. With the removal of the spent
fuel from the SFP, there are no important-to-safety systems,
structures or components required to function or to be monitored. In
addition, there are no remaining credible accidents involving spent
fuel or the SFP that require actions of a Certified Fuel Handler or
Noncertified Fuel Handler to prevent occurrence or to mitigate
consequences. The proposed change to the Design Features section of
the Technical Specifications (TS) clarifies that the spent fuel is
being stored in dry casks within an ISFSI. The probability or
consequences of accidents at the ISFSI are evaluated in the HB ISFSI
Final Safety Analysis Report (FSAR) and are independent of the
accidents evaluated in the HBPP Unit 3 DSAR. Therefore, the proposed
changes will not involve a significant increase in the probability
or consequences of an accident previously evaluated.
(2) Does the change create the possibility of a new or different
kind of accident from any accident evaluated?
Response: No.
The proposed changes reflect the reduced operational risks as a
result of the spent fuel being transferred to dry casks within an
ISFSI. The proposed changes do not modify any systems, structures or
components. The
[[Page 74362]]
plant conditions for which the HBPP Unit 3 DSAR design basis
accidents relating to spent fuel and the SFP have been evaluated are
no longer applicable. The aforementioned proposed changes do not
affect any of the parameters or conditions that could contribute to
the initiation of an accident. Design basis accidents associated
with the dry cask storage of spent fuel are already considered in
the HB ISFSI FSAR. No new accident scenarios are created as a result
of deleting nonapplicable operational and administrative
requirements. Therefore, the proposed changes will not create the
possibility of a new or different kind of accident from those
previously evaluated.
(3) Does the change involve a significant reduction in a margin
of safety?
Response: No.
The proposed changes reflect the reduced operational risks as a
result of the spent fuel being transferred to dry casks within an
ISFSI. The design basis and accident assumptions within the HBPP
Unit 3 DSAR and the TS relating to spent fuel are no longer
applicable. The proposed changes do not affect remaining plant
operations, nor structures, systems, or components supporting
decommissioning activities. In addition, the proposed changes do not
result in a change in initial conditions, system response time, or
in any other parameter affecting the course of a decommissioning
activity accident analysis. Therefore, the proposed changes will not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Jennifer K. Post, Pacific Gas and
Electric Company, 77 Beale Street, B30A, San Francisco, CA.
NRC Branch Chief: Andrew Persinko.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri.
Date of amendment request: October 31, 2007.
Description of amendment request: The amendment would revise
Technical Specification (TS) 3.8.1, ``Essential Service Water System
(ESW),'' and TS 3.8.1, ``AC [Alternating Current] Sources--Operating.''
A note would be added to Condition A, one ESW train inoperable, of TS
3.8.1, and Condition B, one diesel generator (DG) inoperable, of TS
3.8.1 would be revised. The revisions are to allow a one-time
completion time extension from 72 hours to 14 days to support a planned
replacement of ESW piping prior to December 31, 2008, in the licensee's
fall 2008 refueling outage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
[The only change to the plant is that existing ESW piping will
be replaced in the fall 2008 refueling outage. There are no other
changes to the plant and no hardware or equipment will be added to
the plant. This replacement is to address localized degradation of
the ESW piping due to microbiologically induced corrosion.]
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since no
hardware changes are proposed to the protection systems. The same
reactor trip system (RTS) and engineered safety feature actuation
system (ESFAS) instrumentation will continue to be used. The
protection systems will continue to function in a manner consistent
with the plant design basis. The use of polyethylene (PE) piping
[(i.e., replacing existing ESW piping by PE piping)] in the ESW
system in accordance with ASME [American Society of Mechanical
Engineers Boiler and Pressure Vessel Code] Code Case N-755, with
justified materials and design exceptions as noted in [the
licensee's letter dated August 30, 2007 (ULNRC-05434), which
requested relief from the ASME Code to replace the ESW piping by the
PE piping], will [have the PE piping that replaces the ESW piping]
provide an acceptable level of quality and safety. There will be no
changes to the essential service water (ESW) system or [the]
ultimate heat sink (UHS) surveillance and operating limits. [The
licensee's letter dated August 30, 2007,] demonstrates the
acceptability of using the PE piping in this buried ASME Class 3
application [(i.e., replacing existing ESW piping)].
The proposed changes will not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, and configurations of the facility or the manner in
which the plant is operated and maintained. The proposed changes
will not alter or prevent the ability of structures, systems, and
components (SSCs) from performing their intended [safety] functions
to mitigate the consequences of an initiating event within the
assumed acceptance limits.
The proposed changes do not affect the way